ML18143A447

Transmittal of Non-Proprietary Revised Version of Additional Information Re Cycle 8 Fuel Reload

ML18143A447
Person / Time
Site:	Ginna
Issue date:	08/07/1978
From:	White L Rochester Gas & Electric Corp
To:	Ziemann D Office of Nuclear Reactor Regulation
References
Download: ML18143A447 (52)

Text

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+PLEASE RESPOND WITHlN 8 MINUTES gee(/i~/~<

REGUL'ATORY INFORMATION DI TRIBUTION SYSTEM

( RIDS)

DISTRIBUTION FOR INCONING NATERIAL 50-244 REC:

ZIENANN D L ORG:

WHITE L D DOCDATE: 08f07/78 NRC ROCHESTER GAS 8c ELEC DATE RCVD: 08f10/78 DOCTYPL=:

LETTER NOTARIZED:

NO COPIES RECEIVED

SUBJECT:

LTR 1

ENCL 40 FORWARDING NON-PROP COPY OF REVISED VERSION OF Tl-IE ADDL INFO RE Tl-IE CYCLE 8

'FUEL RELOAD AT SUBJECT FACILITY ORIGINAL (SUBMITTED BY LTR DTD 03/27f78.

PLANT NAME: RE GINNA UNIT 1

REVIEWER INITIAL:

XJN DISTRIBUTOR INITIAL:~

~~~~~H:~~~~~~~~H~>>>>~~~

DISTRIBUT1'ON OF TIGRIS MATERIAL IS As FOLLOWS>>>>~>>~>>>>~~>>>>

GENERAL DISTRIBUTION FOR AFTER ISSUANCE OF OPERATING LICENSE.

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FOR ACTION:

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'PLANT SYSTEMS BR>>W/ENCL EFFLUENT TREAT SYS>>M/ENCL EXTERNAL:

LPDR S ROCI.IESTER NY4~'kM/ENCL TERA>>M/ENCL NSI C~~~M/ENCL ACRS CAT B44~W/16 ENCL DISTRIBUTION:

LTR 40 ENCL 39 SIZE:

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Ce REGULtITORY DOCKET,F ROCHESTER GAS AND ELECTRIC CORPORATION o

89 EAST AVENUE, ROCHESTER, N.Y. 14649 10AE LEON D. WHITE, JR.

VICE PRESIOENT IIA1C rl TELEPHONE AREA COOE 7I<<546.2700 August 7, 1978 C.i')

Director of Nuclear Reactor Regulation Attention:

Mr. D..L. Ziemann, Chief Operating Reactor Branch N2 U.S. Nuclear Regulatory Commission Washington, D.C.

20555 Re:

Rochester Gas and Electric Corporation R.

E.

Ginna Nuclear Power Station Docket No. 50-244

'~(~I' Cfl

Dear Mr. Ziemann:

We are enclosing 45 non-proprietary copies of the revised ersion of the additional information regarding the Cycle 8

Fuel Reload at R.

E. Ginna.

Please substitute the enclosed copies for those o 'nally transmitted to you b

letter dated March 27, 1978.

W are a so enclosing en

0) copies o

a ocu n

en z.

ed, t

chment B to Additiona formation Regar the Cycle Fuel Reload" which cont'a s information consi red propriet by Exxon Nuclear Compa y, Inc.

These propr e ary copies are ent to you in confi

nce, and an affi avi request'ng tha they be withheld rom public dis-closure is enclosed.

WQPR(<wR.R.Y Porno&

Qo~mOL( e.4 a-higrgI eu7-~

QQ 1 QQg+g g u Enclosures Very truly yours, r

/

L.

D. White Jr.

~~22OO22i

0 A

Question 1.1 The ENC evaluation model includes a relationship between full rupture pressure and channel flow area blockage.

Provide information to justify that Cycle 8 fuel is bounded by the flow blockage model, indicating the percent flow blockage for the end of Cycle 8 and how it was included in the LOCA analyses.

Response

For the R.E.

Ginna reactor limiting break

case, fuel rod rupture is calculated to occur during the adiabatic refill period; therefore, the effects of blockage occur only in the reflood portion of the transient, and will affect only that portion of reflood with flooding rates below one-inch per second.

For R.E. Ginna, reflood rates are above one-inch per second beyond 300 seconds in the LOCA transient.

As shown by the results given below, the LOCA temperature transient is terminated by 300 seconds.

Therefore, the R.E.

Ginna ECCS analysis is insensitive to. blockage and the worst case in life will be the beginning of life because of higher stored energy.

The calculated fuel clad differential pressure and blockage fraction at rupture are 485.9 psi and 0.28, respectively, for beginning of'life conditions.

Clad rupture is calculated to occur at 37.8 sec at the 5.92 ft location, and the rupture node reached its peak.temperature at 46.2 seconds.

The =peak clad temperature (PCT) node is located at 7.17 ft, and the maximum temperature is reached at 101 seconds.

At 220 seconds, the rupture node and PCT node tem-peratures are 775 and 1476'F, respectively, and the reflood rate is greater than one-inch per second.

At 220 seconds, the liquid level has risen above the core mid-plane (ruptured node),

and the rupture node is quenched.

The blockage versus rupture differential pressure table used in the heatup calculation is given below:

Blocka e Fraction Differential Pressure

( si) 0.75 0.69 0.61 0.26 0.26 0.67

.0.90 0.75 0.58 0.

264.

330.

495.

660.

842.

1172.

1650.

2310.

N:

Question 1.2 Provide a calculated time to collapse for the ENC fuel used in the Cycle 8 reload.

Response

Based on the projected operation for the Cycle 8 reload fuel provided by ENC, the time to collapse is calculated to be greater than 26,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, corresponding to an assembly

'verage burnup of 3S,OOO MWD/MT.

At 26,000

hours, a maximum ovality (ID max.

ID min.)

of 0.066 inches is predicted.

Question 1.3

Response

Describe how the penalties due,to fuel rod bowing for the Exxon fuel were calculated.

Give the numerical values and show how they were applied to LOCA and DNB analyses.

Visual examinations of the Ginna fuel have been conducted over the past five cycles.

These examinations have included both binocular and videotaping.

In all observations at Ginna no fuel rod bowing has been found in fuel assemblies with assembly average burnups in excess of 25,000 MWD/MTU, as might be expected for such fuel design which utilizes stainless steel guide tubes.

Two Exxon Nuclear Company (ENC) fuel assemblies have been in the Ginna reactor for 3 cycles and acquired burnups of approximately 25250 MWD/MTU.

These assemblies have also been videotaped.

The last videotaping occurred after the assemblies had acquired burnups of approximately 15,000 MWD/MTU.

No visual evidence of rod bow was noted at this inspection.

Based on the visual results of no rod bow and the incorporation of stainless steel guide tubes for the ENC reload design',

the rod bow penalties for the Ginna fuel are anticipated to be very near zero.

Since the above rod bow data for Ginna is only qualitative, the rod bow data taken by Exxon Nuclear for its PWR fuel design with zircaloy guide tubes will be used to conservatively bound any rod bow that may occur for the Ginna fuel design with stainless steel guide tubes.

Reference 1 presents data on rod bow measurements taken on ENC reload fuel for the H. B. Robinson reactor utilizing zircaloy guide tubes after one cycle of operation.

Reference 2 presents the results of similar'easurements for the same fuel after two cycles of operation.

The data base obtained from the above measurements include approximately 7000 independent measurements of rod-to-rod spacings for interior as well as peripheral rod bows.

After two cycles of operation, the results indicate rod bow nowhere approaching bow to contact.

1 XN-NF-77-53(P "R. E. Ginna Reload Fuel Design",

December 1977.

(2)

XN-NF-78-28, "H. B. Robinson Fuel Examination Report-February 1978 (To be issued).

The H. B. Robinson rod bow data can be conserva-tively used to evaluate the magnitude of rod bow for the ENC Ginna fuel design because of the following design difference between the Ginna and H. B. Robinson designs:

Ginna use stainless steel guide tubes and 9

grid spacers while H. B. Robinson uses zircaloy guide tubes at 7 grids.

The benefit of stainless steel guide tubes was noted above and the increased number of grids means reduced spacer span and thus less anticipated rod bow.

3 This value includes a 1.2 multiplier to account for cold-to-hot variations in measured rod spacings and a 0.844 multiplier to account for the smaller grid spacer spacing in the Ginna fuel (22.1" vs. 26.19" for H. B. Robinson).

The calculation of the DNB rod bow contact penalty is based on DNB tests with rod bow as referred to in the NRC's Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Cal-culations for Light Water Reactors.

The fractional reduction in MDNBR was calculated for the three most limiting condition~]) and III transients as reported in ZN-NF-77-40 Transient Btu r-ft Pressure sia

• 2-pump coastdown 290,102 2234 Control rod withdrawal 328,331 2279 Loss of load 305,971 2437 As the contact penalty increases with pressure, an addition 60 psi was added to the reported transient pressure in the calculation to account for measurement uncertainty.

The two expressions for bow-to-contact DNB penalty were evaluated at the average hot rod heat flux from each transient.

Pressure was linearly interpolated from the results.

At e point zn tame w zc the transient MDNBR occures.

(3) F. J. Markowski and J.

D. Kahn, "Plant Transient Analysis for the R. E. Ginna Unit 1 Nuclear Power Plant," XN-NF-77-40, November 1977.

These transients envelope all the DNB limiting Condition II and III transients reported in ZN-NF-77-40. It should be noted that the transient analysis was performed using a nuclear peaking value significantly higher than the maximum allowed nuclear peaking of 2.32 as specified by the Plant Technical Specifications.

The results of the analysis are given below and indicate that the MDNBR values for the three most limiting transients are well above a limiting 95/95 condifence of probability statement of DNB.

Therefore, no DNBR rod bow penalties are required for Ginna.

TABLE I CALCULATED TRANSIENT NDNBR VALUES FOR THE GINNA REACTOR Transient 2-pump coastdown Control rod withdrawal Loss of load MDNBR With Rod Bow Penalt

The reported transient analysis was based upon a

hot channel model for the ENC reload fuel.

An updated steady-state thermal-hydraulic calculation performed with subsequently measured hydraulic, loss coefficients for the ENC reload fuel shows a

2% additional reduction in flow to the hot ENC channel than previously predicted.

(his causes the MDNBR at 118% overpower and at F

= 2.80 to be reduced from 1.47 to 1.44 for theqENC fuel.

Applying this ratio to the most limiting transient results (2-pump coastdown) reduces the MDNBR with rod bow to C

3 with a total nuclear peaking factor well above current license limits.

This value of MDNBR is well above a limiting 95/95 confidence of probability statement of DNBR value.

The R. E. Ginna Plant Technical Specifications incorporate a total nuclear peaking augmentation factor of 1.082 in calculation of the ECCS safety limits.

This factor is adequate to accommodate nuclear augmentation due to rod bow in a limiting assenbly any time during the anticipated life of the reload fuel in the reactor'ore.

Therefore, no additional penalty due to rod bow needs to be applied to calculation of LOCA limits.

The combination of nuclear factors is accomplished as follows:

F

= 1.0+/+

g

+g

+g q

where, F

= Total nuclear peaking augmen-tation factor due to uncertainties

= Fractional augmentation due to Rensification

= Fractional augmentation due to 5ngineering (pellet density,

diameter, enrichment)

= Nuclear measurement and calculation uncertainty SB = Nuclear augmentation due to rod bow

The latter three factors are statistically indepen-dent while fuel pellet densification is expected to occur almost uniformly throughout the reactor core.

The following table lists the bow augmenta-tion factor as a function of burnup based upon ENC's measurement of rod bow of ENC supplied 15 x 15 zircaloy guide tube fuel (XN-NF-77-52).

The interpretation of bow data is conservative and is in accord with Staff-imposed guidance.

The table also lists the required total augmentation factor to protect all uncertainties.

EXPOSURE DEPENDENT TOTAL NUCLEAR PEAKING AUGMENTATION APPLIED TO EXXON NUCLEAR FUEL Exposure

~Cele Region Average Expected EOg Burnup, 10 NWD Fractional Augmentation Due to Bow Total Nuclear Peaking Augumentation Factor 6.7 17.0 27.3 33.8

.014

.032

.046

.054 1.066 1.073 1.081 1.086

Thus, even if bow were assumed to occur in fuel supplied for Ginna, an assumption which is not justified based on the evidence, no penalties need be assessed to provide for margin to limits.

Question 2.1 Provide a detailed description of how the individual incore power distribution co-efficients were calculated.

What are the uncertainties of the calculations?

Response

The individual incore power distribution coefficients are the two dimensional PDQ calculated core assembly and pin relative powers.

The method of calculation of core power distributions is provided in Document XN-75-27, "Exxon Nuclear Neutronic Design Methods for Pressurized Water Reactors,"

June

1975, by F. B. Skogen and its two supplements issued in September 1976 and December 1977.

A 5% uncertainty is assigned to the predictions.

This value is supported by previously measured power distributions in similar plants.

Question

2. 2

Response

Provide the predicted axial distribution for the peaking factor F N.

Q The predicted axial F distributions at BOC and N

EOC conditions are sh8wn in Figure 1.

Both distributions are at, equilibrium xenon conditions.

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Question 2.3 List the uncertainties in determining the required shutdown margins for the Cycle 8

reload.

Response

The required shutdown margin is the amount of reactivity by which.the core must be subcritical, at HZP, ARI minus the most reactive rod conditions, to prevent a return to critical following a small steam line break and to assure CHF is not reached during the return to power following a large steam break.

A bounding value is assigned to the required shutdown margin by performing the steam line break analyses in a conserva-tive manner (see XN-NF-77-40).

Additionally, the predicted rod worth which must provide the required shutdown=margin is reduced by l0% in the determination of whether the required shutdown margin is met.

Question 3.1

Response

Describe in detail the design difference between

~

the ENC and Westinghouse fuel assemblies which have impact on their thermal hydraulic performance.

Give the numerical value of the hydraulic loss coefficient for the ENC assemblies and compare it to the loss coefficients for Westinghouse fuel.

What is the difference in the flow areas in the ENC and Westinghouse fuel assemblies.

The design differences between ENC and Westinghouse fuel assemblies which have an impact on their thermal hydraulic performance are illustrated in the table below.

The impact of the geometrical difference on the fuel's thermal-hydraulic per-formance is very small.

MECHANICAL DESIGN DIFFERENCES BETWEEN ENC AND WESTINGHOUSE FUEL ASSEMBLIES IMPACTING ON THERMAL HYDRAULIC PERFORMANCE Design Factor ENC Fuel Westin house Fuel Fuel Pellet Diameter (in.)

Enrichment Clad ID (in.)

Clad OD (in.)

Rod to Rod Spacing (in.)

Control Rod OD (in.)

Bare Rod Flow Area (in.

)

2 Wetted Perimeter (in.)

Heated Perimeter (in.)

0.3565 0.0310 0.364 0.424

0. 132

0. 540

31. 81 266.9 238.4 0.3659 0.0310 0.3734 0.422 0.134 0.5375 32.08 265.7 237.3 Single phase isothermal hydraulic tests of both Westinghouse and the ENC reload fuel designs were conducted in ENC test facilities.

The same test internals hardware was utilized in the testing of each fuel type.

The purpose of the testing was to" determine the component hydraulic coefficients of the lower tie plate (including lower core support plate),

spacers, bare rod (friction) and upper tie plate (including plugging device and upper core support plate).

In conjunction with similar data obtained from earlier tests on the Westinghouse fuel this information can be used as.a basis for hydraulic

.compatibility determination for a mixed core of ENC and Westinghouse fuel.

12

Differential pressure measurements were made at 14 Reynolds numbers (including replications) between 100,000 to 300,000.

These data were used to deter-mine empirical relationships describing the single-phase loss coefficients of the Exxon Nuclear Ginna XN-1 fuel assembly and its components'he resulting single-phase loss coefficients were enveloped by the design uncertainties used in the thermal-hydraulic core analysis calcu)g)iona reported in the Ginna licensing submittal The empirical relationships describing the ENC assembles single-phase loss coefficients and a

diagrammatic comparison of the ENC and Westinghouse assemblies loss coefficients are presented in Attachment A to these questions and responses.

(1) Experimental Hydraulic Evaluation.of Ginna Reactor Fuel Assembly

("Westinghouse"

) Smith 6 Greer, Battelle Northwest, Feb.

1975 (2) XN-NF-77-52 R.E.

Ginna Reloads Fuel Design, December 1977.

13

Question 3.2 Justify a 15 percent uncertainty used in cal-culating the hydraulic loss coefficients.

Response

For purposes of preliminary design evaluation, an uncertainty of + 15 percent was applied to the calculated component hydraulic loss coefficients.

These components include upper and lower nozzles and spacers.

No uncertainty was applied to the bare rod friction losses.

As added conservatism in the evaluation, a +

5 percent uncertainty was applied to the reported test results (nozzles and spacers) for the Westinghouse R.

E. Ginna fuel.

On an assembly overall basis (including" friction),

these uncertainties are

+ 12 percent and

+

4 percent, respectively.

Subsequent to submittal of XN-NF-77-53, Exxon Nuclear Company has completed single phase isothermal hydraulic testing of the ENC/R.E.

Ginna proof-of-fabrication fuel assembly.

The results of this

testing, together with a summary of the results of a test earlier performed on the Westinghouse R.E.

Ginna fuel (both tests in ENC test hardware and ENC facilities), are presented in the response to ques-tion 3.1.

Both tests were conducted with the same test internals hardware (core plates,

channel, etc.)

and with the same pressure taps.

Facility instru-mentation was calibrated with identical standards referenced to NBS secondary standards.

Therefore, the mean value of assembly hydraulic coefficients was used to make final appraisal of.hydraulic com-patibility of the two fuel types.

The results of the testing show the hydraulic loss coefficient of the two fuel types to be similar.

For two fuel assemblies, one of each type operating at identical power, the flow difference would be approximately 3 percent.

The results of analysis of three cases appraising hydraulic compatibility are presented in Table 3.2.

'Case I is as was presented in XN-NF-77-53, and is based upon the previously described calculated hy-draulic coefficients for ENC fuel, including un-certainties.

Cases II and III were analyzed using the test data reported in the response to question 3.1.

Case II considered a mixed core comprised of~26 percent ENC fuel and balance W fuel.

Cases I and II were performed using standard ENC metho-dology including:

1) a core-wide flow distribution analysis to determine limiting assembly flow rate; and 2) an assembly subchannel analysis using the assembly flow rate from step l.

14

The measured loss coefficient of the ENC fuel is slightly higher than for the W fuel.

When the core is comprised of 26 percent ENC fuel (Cycle 8), the ENC fuel will be located in the core periphery and will not be limiting; hence, the fuel in the center of the core with higher radial peaking will receive more flow than with an all W fuel core.

(Case I assumed

[prior to test data]

the ENC fuel hydraulic resistance to be low and thus the center region fuel would receive less flow.)

Because of this and other conservatisms applied in the Case I analysis, Case I exhibits the MDNBR of the three cases.

Case II (using test data) analyzed a similar mixed core arrangement but conservatively assigned highest radial peaking (Fr= 1.60) to ENC fuel even though situated on the periphery of the core.

This case is less limiting than the prior documented Case I.

Case II analyzed a mixed core as would be possible in Cycle 9 wherein the ENC fuel will be situated in the core central region and therefore could experience maximum radial peaking.

Of the three

cases, this is most realistic.

The analysis was performed in two,steps as before except the sub-channel analysis was performed with two adjacent fuel assemblies, one of each manufacture.

This analysis fully reflected the changed axial pressure distribution and inter-assembly flow between the fuel assemblies induced by the different measured nozzle and spacer loss coefficients.

The analysis was also conservatively performed with the W

fuel radially peaked at factor of 1.06 so that this cooler assembly absorbed maximum flow.

This case is seen to be the least limiting of the three.

Therefore, Exxon Nuclear Company concludes the two fuel types to be thermal hydraulically com-patible and further concludes that introduction of this fuel in the reactor core in the amounts proposed (from roughly 1/3 core to full core) will not result in reduction of DNBR margins below license limits.

15

TABLE 3.2

SUMMARY

OF HYDRAULIC COMPATIBILITYFINDINGS*

CASE III Average

~Ass flow, 10 6

.536

.536

. 536 Limiting lb/hr Ass flow, 10 lb/hr 6

.510

.511

.517 Assy radial factor

1. 60

1. 60

1. 60 Percent of ENC fuel zn core MDNBR 1.43 1.44 1.50

• Analysis performed at 118 percent steady state core power, total nuclear peaking 2.80.

16

Question 4.1:

Provide the magnitude of the local reactivity, assembly

power, and axially integrated rod power errors associated with the worst fuel assembly misloading which may occur and which would not be detected by the core instrument system.

Response

The Ginna Cycle 8 core is expected to have 78% margin to the FQ value used in the DNB analysis, 47K to the F

Technical Specification limit and 25%

margin to the F< limit as indicated in the Ginna Cycle 5

SAR (XN-NF-77-53).

In addition, all accident analyses show considerable margin to design limits.

A conservative estimate of the minimum power difference that would be detected is 15%.

Thus, the Ginna Cycle 8 core will have margins to limits that signifi-cantly exceed undetected assembly power differences which could occur from a mis-loading error.

This general subject is also under generic review by our fuel sup-plier, Exxon Nuclear Company.

17

I ~

Question

4. 2 Describe the statistical analysis which was per-formed in order to determine that less than one percent of the fuel rods in the Ginna plan will experience DNB during the postulated locked pump rotor accident where DNBR reaches the minimum value of 1.23.

Response

The statistical analysis to determine the number of rods which will experience DNB during the locked rotor transient was performed as follows:

(1)

The number of rods experiencing departure from nucleate boiling ratios (DNBRs) in the range from 1.5 down to the minimum value of 1.23 was determined assuming the data to be normally distributed; (2)

The number of rods having a

DNBR within this range was then multiplied by the probability that a rod having this DNBR would experience DNB; and (3) This joint probability was then summed over the core and divided by the total number of rods in the core to give the expectation of DNB.

The number of rods in a DNBR interval multiplied by the maximum probability for DNB in this interval and summed over the core (joint, probability) gives expecta-tion of 0.033%

(or about 7 rods) experiencing DNB, which is well below 1% of the fuel rods as stated.

18

Question 4.3

Response

In the analysis of the rod withdrawal transient, it is stated that for partial power loading the margin to DNB is slightly decreased in the over-temperature aT regime.

Indicate by how much this margin is reduced.

For a rod withdrawal transient from 80% power, the DNBR is reduced by 0.01 compared to a tran-sient from full power; at 60%, the reduction is 0.04.

Since at 102% power the MDNBR was predicted to be 1.73, the minimum DNBR which would be pre-dicted during a rod withdrawal from 60% power is about 1.69.

19

Question 4.4

Response

Explain your statement that. in the case of the rod withdrawal transient, the change in plant response is mainly dependent on the plant pro-tection system.

The trip setpoints associated with the rod with-drawal transients remain unchanged from the yyl ly'., l ~d of the plant. response (reactor trip and associated core conditions at the time of trip) will be principally a function of the setpoint values.

20

Question 4.5

Response

In the plant transient analysis for the Cycle 8

reload, only the most limiting transients for the existing fuel were reanalyzed.

These transients were considered limiting because they were shown in the reference cycle analysis to have the least margin to technical specification limits.

The other transients, which showed larger margin in the reference cycle analysis, were not analyzed.

Describe your criteria for making this choice.

Give quantitative examples.

Reload fuel is designed with hydraulic, neutronic, and mechanical compatability as a primary basis.

It then follows there is no significant effect, of the reload fuel design on established reference cycle transients.

This is verified by reanalyz-ing those transients which were determined to be potentially limiting in the reference cycle analysis.

The potentially limiting transients are those which may approach the plant technical specifi-cation limits on:

(1)

MDNBR (loss of flow);

and/or (2) peak system pressure (loss of load).

In addition, analyses are presented to ensure:

(1) the adequacy of current.trip setpoints in maintaining margin to DNB (specifically the rod withdrawal transient);

and (2) the adequacy of the predicted EOC shutdown margin in protecting core integrity (steam line break).

Other reactivity insertion transients of less severity, such as loss of feedwater, startup of inactive

loop, and excessive load, which showed a greater margin to DNB in the reference cycle analysis, are not reanalyzed.

A comparison of the MDNBR reported in the reference cycle analysis for the MDNBR limiting and two non-limiting transients is provided below:

Technical Supplement Accompanying A lication To Increase Power

\\

Loss of Flow Excessive Load Loss of Feedwater 1.49 1.66 2.00 A large DNB margin is seen to exist between the non-'limiting and MDNBR limiting transients.

The analysis of the non-limiting (with respect to DNB) transients in the reference cycle provides assurance that the plant safety and protection systems (trip setpoints, auxiliary feedwater, 21

etc. ) are adequate to protect against an approach to DNB.

As the setpoints of these protect,ion systems remain unchanged, and it has been demon-strated that, the reload fuel has adequate margin to DNB during more severe reactivity insertion transients (rod withdrawals),

a larger margin of protection to DNB would be evident for the remain-ing transients.

Review of the system responses for the analyzed transients also confirms that the reload fuel has not significantly altered the system response.

Thus, by reanalyzing those transients that were previously determined as limiting ensures that the most restrictive operat-ing conditions are identified.

22

Question 5.1 Justify, by giving the appropriate references, why the small break LOCA analysis was not per-formed by ENC for Ginna reload Cycle 8.

Response

A large break spectrum was analyzed for the R.

E. Ginna reactor by the NSSS vendor.

Exxon Nuclear Company also performed a large break analysis for the R. E. Ginna reactor.

The re-sults for the large break analysis are compared in the table below.

The table illustrates that.

the Peak Clad Temperature (PCT) versus break size trends from both calculations are similar and the PCT for the most. limiting break is in close agreement.

Exxon Nuclear Company has performed small break analyses for the H. B. Robinson 3-loop Westing-house reactor and compared the results with th'e 3 loop sensitivity analysis performed by the NSSS vendor.

These results are compared in the table below.

The table shows that small break PCT's are well below the 2200'F limit.

The break spectra results also indicate similar PCT versus, break size trends.

A small break analysis was performed for the R. E. Ginna reactor by the NSSS vendor.

The results of this analysis are presented in the table.

Since the ENC WREM-II small break analy-sis model remains unchanged from that used in the H. B. Robinson analysis, an Exxon Nuclear Company analysis of small breaks for R. E. Ginna would give results similar to those of the NSSS vendor and previous 3-loop plant results.

These results are substantially below the limiting large break results and clearly within the require-ments of 10CFR50.46.

Lar e Break DECLG PCT'F D

C

= 1.0 CD = 0.6 CD = 0.4 R.E. Ginna (by NSSS vendor)

R.E. Ginna (by NSSS vendor)

R.E. Ginna (by ENC) 1757 1805 1747 1853 1898 1957*

1922

• Reanalysis to account for 10% Steam Generator tubes plugged and upper plentum temperature equal to TH.

Sensitivity analysis showed PCT vs. size trend remains unchanged.

23

Small Break size-ft PCT'F 1.0 2

0.5 0.349 0.196 0.097 0.049 0.022 westinghouse 3 Loop Plant (by NSSS vendor)(NCAP-8356) 1457 1546 1435 1703 1286 1442

( 800 H.B. Robinson (by ENC) 1038 1457 R.E.

Ginna (by NSSS vendor) 1443 1407 1469 1668 1167 24

Question 5.2 For the most limiting large break LOCA analyses include the following additional plots:

(a)

Energy release to the containment vs.

time (b)

Fluid quality, mass velocity and temperature at hot spot location vs.

time (c)

Core pressure drop (lower plenum to upper plenum) vs. time.

Response

The energy release to the containment versus time for the 0.4 DECLG break is given in Figure 5.2-1.

The fluid quality versus time for the center volume in the hot assembly for the 0.4 DECLG break is given in Figure 5.2-2.

The mass flux versus time for the inlet and outlet junctions to the center volume in the hot assembly for the 0.4 DECLG break is given in Figures 5.2-3 and 5.2-4, respectively.

The hot spot fluid temperature versus time for the 0.4 DECLG break is given in Figure 5.2-5.

The core pressure drop (lower-plenum-to-upper-plenum) versus time for the 0.4 DECLG break is given in Figure 5.2-6.

25

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FIGURE 5. 2-6 CORE PRESSURE OROP VERSUS TINE 28 32 36 40

Question 6.1 Startup physics tests selected from Reg.

Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors,"

should be performed.

The purpose of the startup test program is to (1) verify that the core was cor-rectly loaded and there are no anomalies present which could cause problems later in the cycle, and (2) verify that the calculational model that has been used will correctly predict core behavior dur-ing this cycle.

A test abstract summarizing the test objective, test method, and acceptance criteria should be presented.

Specifically the following questions are submitted on the test programs:

Response

The objective of the Startup Physics Test program for Ginna Cycle 8 is to reconfirm that the core is correctly loaded, to ensure that there are no anomalies present which could cause'roblems later in the cycle and to verify the calculational model that was used correctly predicts core behavior.

The test methods and acceptance criteria are pre-sented in the following responses.

This information supplements the description previously supplied in Attachment B to Application to Amend Provisional Operating License DPR-18 dated January 6,

1978.

32

Question 6.l.a Describe in detail the tests being done to check for a misloaded assembly?

What assurances are there that the core is as expected before going to powers

>5% rated powers

Response

Prior to any movement of fuel assemblies a de-tailed, step-by-step fuel loading procedure is prepared.

This procedure specifies the step-by-step movement of each fuel assembly through-out the fuel handling.

The resulting loading pattern is checked against the core loading pat-tern supplied by the fuel supplier to assure that there will be no misloaded fuel assemblies.

This procedure is then followed to load the core.

Prior to placing a fuel assembly into a core loca-tion the fuel assembly identification number is read with an underwater TV and verified to be the assembly specified in the procedure.

Following core loading the entire core is mapped using an underwater TV camera.

The fuel assembly identification numbe'rs and core location are re-corded and then compared against the core loading pattern to assure that there are no misloaded assemblies.

As a final independent

check, on all rods out, full

core, nominal zero power flux map is taken and compared against predicted values to insure there are no misloaded fuel assemblies.

The flux map is taken at a power level less than 5% power.

There-

fore, by the checks mentioned above the core is as expected before going to powers

>5% rated power.

33

Question 6.l.b Describe the procedures for the control rod-drive tests and droptime tests.

Include the acceptance criteria and the procedures to be followed if the acceptance criteria are not met.

Response

The control rod droptime tests are done at hot full flow conditions.

A selected bank of control rods is fully withdrawn.

An oscillograph is then connected to the rod position detector primary coil and the stationary gripper coil of the rod to be tested.

The control rod is then dropped into the core.

From the data recorded on the oscil-

lograph, the start of rod motion can be determined, the time at which the rod enters the dash pot can be determined, and the time at which the rod bottoms can be determined.

The acceptance criterion for this test is that Technical Specification 3.10.3.1 be met.

If this criterion is not met, Technical Specification 3.10.4 will be imposed and the cause of the problem will be determined and corrected.

The control rod drop-.

time test will then be repeated for the rod/rods that did not meet the acceptance criterion.

Throughout the cycle, control rod-drive tests a'e preformed periodically as required by Technical Specification 4.l.b.

For this test the control rods are moved in and out a predetermined number of steps to insure control rod operability.

34

Question 6.l.c Provide the details of the procedures for the criti-cal boron concentration tests.

Discuss how correc-tions are made to the measured data and how the measured data is compared to the predictions.

What is the acceptance criteria and what are the proce-dures if the acceptance criteria is not met.

Response

The critical boron concentrations will be measured for All Rod Out (ARO)g D bank in, and D +

C bank in conditions.

For the ARO test, the critical boron concentration will be measured with the D

bank slightly inserted.

Then the D bank will be fully withdrawn and the resulting reactivity will be measured with the reactivity computer.

Since the boron worth is known for the ARO condition, the reactivity measured can be converted into an equiva-lent ppm of boron.

This is added to the critical boron concentration measured with D bank slightly inserted to obtain the ARO critical boron concentra-tion.

The same methods are used to obtain the D bank in and the D +

C bank in critical boron concentrations.

The average critical boron concentrations obtained by the above method for a given configuration will be compared to the predicted concentration for that con-figuration.

The acceptance criterion will be +75 ppm.

If this acceptance criterion is not met, the measurement data will be reviewed and the fuel sup-plier will be asked to review his predictions.

Concurrent with this review the remainder of the zero power physics testing will be completed to see if any other differences are present that could aid in determining the cause of not meeting the accep-tance criteria.

If after the above actions, the acceptance criterion still can not be met an evalua-tion will be performed on the effect of this dif-ference on parameters used in the accident analysis.

If the accident analysis is unaffected by this dif-

ference, the core will be allowed to go above 5%

rated power.

35

g ~

Question 6.l.d Describe in detail the procedures and methods used for the temperature reactivity coefficient tests.

Also provide the acceptance criterion and the procedures to be followed if the acceptance criteria is not met.

Response

The isothermal temperature coefficient will be measured for the

ARO, D bank in and D +

C bank in core configurations.

'I For the ARO i othermal temperature coefficient measurement the core will be made critical with D

bank slightly inserted.

Using the secondary side steam dump control, the reactor coolant system will be heated up and cooled down at approximately l0 F/Hr.

A reactivity computer is employed to record the reactivity change induced by the change in primary system temperature.

The isothermal temperature coefficient is determined by obtaining the ratio of the change in reactivity to the change in tem-perature.

The same methods will be used to obtain the D bank in and D +

C banks in isothermal temperature coef-ficient.

A moderator temperature coefficient will be obtained by subtracting the calculated doppler coefficient from the measured isothermal coefficient.

The acceptance criterion for these tests is that the moderator temperature coefficient is non posi-tive. If the criterion is not met, appropriate operating restrictions, such as rod withdrawal limits, will be developed so that the plant will operate with a non-positive moderator temperature coefficient.

36

Question 6.1.e Provide the details of the regulating control rod group reactivity worth tests.

Give the predicted worth of each group to be measured, the stuck rod worth and the predicted total worth for all rods.

What is the acceptance criteria for these tests?

What are the procedures if the acceptance criteria is not met?

Response

The differential worth of banks D,

C, B,

and A will be measured.

From these measurements the integral bank worths can be obtained.

The differential bank worth is measured by first obtaining a critical configuration with the bank of interest approximately full withdrawn.

Then boron dilution is initiated.

The bank of interest is then inserted a given number of steps with the reactivity computer monitoring flux and reactivity response.

The reactivity response calculated by the reactivity computer is then recorded for the associated change in bank position.

This sequence is repeated until the bank is fully inserted.

From the resulting data a plot of differential bank worth versus bank position is developed.

The integral bank worth will be developed from these data.

The acceptance criteria for these tests are that the measured individual bank integral worth be within 15% of the predicted values and that the total worth of all four banks be within 10% of the predicted value.

If the criterion on individual bank worth is not met, an evaluation will be per-formed to determine the cause and any potential impacts.

If the criterion in total worth of the four banks is not met, additional banks will be measured until the measurement is within 10% of the prediction.

This will be continued, if necessary, to measurement at an N-1 rod inserted condition.

The result of an N-1 measurement, with appropriate allowance for measurement uncertainty, will be compared to the value assumed in the Safety Analyses.

The predicted values for the regulating control rod group reactivity worth tests have not yet. been calculated but will be presented, as required, in the Summary Report of the Physics Startup Tests.

37

Question 6.1. f Provide the details for the core power distribution

tests, including at what power level the tests are performed.

Describe in detail the methods used to predict the assembly by assembly power as well as the analyses of the data obtained during the measure-ments.

(What is the assembly by assembly acceptance criteria?

How are tilts accounted for in the analysis of the data?

If a 1/4 or 1/8 core map is the result of the measurement, what method is used to determine the as embly power for an assembly having its symmetric assemblies instrumented?

Response

Full core flux maps will be taken for the ARO and D bank in configuration prior to exceeding 5%'ated power.

The maps will be used to compare detector data with predictions for the actual detector loca-tions.

If possible, these flux maps will be reduced using the INCORE computer program;

however, hand reduction is acceptable if the INCORE program can not be run.

In compliance with Technical Specifi-cation 3.10.2.1 a full core flux map will be taken and reduced with the INCORE program prior to ex-ceeding 50% rated power to insure that design limits are not exceeded.

The methods used to predict the assembly by assembly power are described in the response to question 2.1.

The acceptance criterion is that the plant Technical Specification on peaking factors be met.

As an aid in evaluating the power distribution maps, the dif-ferences between measured and predicted assembly power levels are reviewed.

General criteria for these comparisons are that the difference for assemblies with relative power (1.0 be less than 15% and the difference for assemblies with relative power 1.0 be less than 10%. If these differences are ex-

ceeded, an evaluation shall be performed to ensure that the peaking factor Technical Specifications will be met. If such an evaluation indicates that the peaking factor limits will be met, no further action is necessary.

Only full core flux maps will be taken during the Startup Physics Test Program, neither quarter nor eighth core maps will be taken.

Tilts will be calculated by the INCORE program and compared to Technical Specification limits.

38

I E ~

~

gl ~

Question 6.1.g Provide details of the power coefficient measurement near full power.

What methods are used to compare measured value with predictions?

What is the accep-tance criteria for this test and what procedures are followed if acceptance criteria is not met.

Response

The power coefficient measurement is done at a power level greater than 65% after the fuel preconditioning requirements have been satisfied.

A power oscillation is created by ramping the turbine approximately

+10%

and

-10% rated power.

The core T

is then kept equal to T by inserting and wi8K8rawing control rods.

The Reactivity computer is used to obtain the reactivity associated with each control rod move.

From these data the power coefficient can be cal-culated.

If the difference between the measured and calculated power coefficient is within 30%, the results of this test will be considered acceptable.

If this cri-teria can not be met the test will be repeated and if the average measured value is within the envelope of power coefficients used in the accident analysis, there is no safety concern and the results of these tests will be considered acceptable.

If the average value of the tests is outside the envelope of power coefficients used in'he accident

analysis, operating restrictions will be imposed which will result in safe operation of the reactor until the accident analysis is re-evaluated using the measured data and safe operation of the cycle can be shown.

39

Question 6.1.h Discuss the changes to be made to the plant's com-puter prior to cycle 8 operation.

This should describe:

1)

What elements changes from cycle 7 to cycle 8; coefficients, constants, correlations etc.

and why they change.

2)

What codes and methods are used to establish the new values.

3)

What quality assurance procedures (testing) are used at the site to verify that changes have been correctly made.

Response

No changes will be made to the plant computer for cycle 8 operation.

The only element that changes from cycle 7 to cycle 8

is the source terms used by the INCORE computer code.

These terms are generated by the fuel supplier as described in the response to question 6.l.f.

The source terms are then supplied to Rochester Gas and Electric Corp. in the form of punched cards.

These cards are then used to create an UPDATE file on a comercial CDC computer system.

The UPDATE file is printed out and the printout is compared against power dis-tributions supplied by the fuel supplier.

Another check is performed by running the INCORE code to obtain the predicted power distribution (source terms).

The predicted power distribution is then compared to that supplied by the fuel supplier.

Both of these checks are done when source terms are changed to ensure that the change has been correctly made.

40

Question 6.2 A summary report of the physics startup tests should be submitted to NRC within 90 days following comple-tion of the startup test program.

This report should include both measured and predicted values.

If the difference between the measured and predicted value exceeded the acceptance criterion, the report should discuss the actions that were taken and justify the adequacy of these actions.

An outline which may be used for the physics startup tests summary report is included.

Please state that such a report will be provided.

Response

In accordance with Technical Specification 6.9.1, a summary report of the startup physics testing will be provided within 90 days following com-,

pletion of the startup test program.

It will in-clude measured and predicted values.

If the test acceptance criteria are not met, the report will describe actions taken and the adequacy of such actions.

4 8

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I 13