ML20126C851

Safety Evaluation Supporting Amend 29,to License DPR-54

ML20126C851
Person / Time
Site:	Rancho Seco
Issue date:	03/31/1980
From:	Office of Nuclear Reactor Regulation
To:
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ML20126C846	List: ML20126C843 ML20126C847 ML20126C851 ML20126C856
References
NUDOCS 8004100251
Download: ML20126C851 (13)

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9 O g/"%o, UNITED STATES l I, .. J[ } NUCLEAR REGULATORY COMMISSION l

WASHINGTON. D. C. 20555

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l SAFETY EVALUATION BY THE OFFICE OF I

..... NUCLEAR REACTOR REGULATION SUPPORTING j AMEflDtiENT NO. 29 TO LICEllSE NO. DPR-54 l SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SEC0 NUCLEAR GENERATING STATION 1 DOCKET NO. 50-312 l l

1.0 Introduction By letters dated October 3,1979, as supplemented by letters dated February 19 and 27, 1980, (References 1, 2, and 3 respectively),

Sacramento Municipal Utility District (SMUD) requested amendment of Appendix A to Facility Operating License No. DPR-54 for Rancho Seco Nuclear Generating Station. Section 6 summarizes the proposed changes of this amendment to the Technical Specifications (TS).

The SHJD submittal of October 3,1979 was presented to support operation of Cycle 4 following the refueling performed at the end of Cycle 3. As such, the analysis presented in the submittal was i I

based on the intended exposure for Cycle 4 of 335' effective full pcwer days (EFPD),, Information submitted describes the fuel system design, nut! ear design, thermal-hydraulic design, accident 1

, analyses and startup test program.

The refueling of Rancho Seco for Cycle 4 will result in a core loading consisting of 52 (batch 6) fresh Mark-84 assenblies with an initial enrichment of 3.21 wt 'a 235U ,13 (batch 1 B) and 56 (batch 5) once burned assemblies with initial enrichments of 2.01 and 3.04 wt : 235U respectively, and 56 (batch 4) twice burned assemblies with an initial enrienment of 3.19 wt : 235U.

The fuel management has been changed from a conventional three fuel batch out-in scheme to a four fuel batch in-out-in scheme.

The key feature of this type of fuel management is the extensive use of fixed burnable poison in fresh reload' fuel which will be .

loaded in the core interior rather than on the core periphery. '

The maximum fuel batch exposure at the end of Cycle 4 is pre-dicted to be 32,300 MWD /MTU and hence is less than the value of 33,000 MWD /MTV used in the staff environmental considerations.

This report addresses Cycle 4 operation only. Operation of successive extended (nominal 18 month) fuel cycles will result in fuel batch exposures in excess of 33,000 MWD /MTV during Cycle 6. The environmental impact of extended fuel burnup is to be addresed prior to Cycle 6 operation.

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2.0 Evaluation of Mocification to Core Design 2.1 Fuel System Design The 52 fresh Mark B-4 fuel assemblies which are to be loaded for l Cycle 4 are mechanically identical to previously approved and utilized fuel assemblies at Rancho Seco and other Babcock and Wilcox (B&W) supplied nuclear steam supply systems (NSSS).

The fuel pellet end configuration has changed from a spherical dish for batches I through 5 to a truncated cone dish for batch 6.

The new design reduces pellet end laminations during manufacturing while maintaining the same end void volume. Fuel performance will not be adversely affected by this change.

Modified Burnable Poison Rod Assemoly (BPRA) retainers (Reference a) are to be used in Cycle 4 to insure positive retention of the BPRAs.

These retainers have been previously aoproved for retention of Orifice Rod Assemblies (ORAs). Mechanical and thermal-hydraulic l coroatibility of tne BPRA retainers has been previously reviewed and found acceptable (Reference 16). l l

2.1.1 Claccing Creep Collapse Fuel rod cladding collapse analyses have been performed for the most limiting (i.e. , twice exposed batch 4 fuel assemblies) fuel assembly to be used in Cycle 4. The analyses were performed according to the methods and assumptions cescribed in References 5 and 6 These analyses predict that the time to rod cladding collapse will be in excess of 30,000 ef fective full power hours (EFPH). The maximum batch 4 assembly burnup during Cycle 4 is predicted to be 21,144 EFPH (Table 4.1, Reference 1). We concluded that cladding collapse has been adequately considered. ,

2.1.2 Cladding Stress and Strain j Stress calculations have been performed for a generic fuel rod model and strain calculations for a generic pellet model. These models and calculations have been approved for prior Rancho Seco reloads. The licensee has asserted that Cycle 4 parameters are enveloped by these generic models. The licensee's calculations sh> that in all cases the margin is in excess of 30% of the maximum specified unirradiated yield strength.

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2.1.3 Fuel Thermal Desian The addition of the batch 6 fuel does not introduce significant differences in fubl thermal performance relative to the other fuel remaining in the core. The predicted linear heat rate (LHR) to centerline melt (20.4 kW/f t) is the same for batches 4, 5 and

6. LHR capabilities are based on centerline fuel melt and were established using the approved computer code TAFY-3 (Reference 6) with consideration for fuel densification. At the core average LHR (6.27 kW/ft), the licensee predicted nominal fuel temperatures of batches 4, 5 and 6 would be approximately 1350*F. These values are typical of all PWRs.

Based on the Cycle 4 predicted values and current approval of the analytic techniques used to make these predictions, the staff considers the fuel thermal design acceptable.

l 2.2 Nuclear Desicn Figure 3-l' (Core Leading Diagram-Rancho Seco, Cycle 4) of l Reference 1 indicates the core loading arrangement for Rancho Seco, Cycle 4; the initial enrichments and burnup distr'i-butions are given in Figure 3-2 (" Rancho Seco BOC 4 Enrichment ,

and Burnup Distribution"). An unconventional fuel management j scheme has been utilized.

An in-out-in fuel management methodology has been adopted. Fresh - l (batch 6) 3.21 wt;, 2350 fuel will be loaded in the core interier l in a checkerboard pattern. Next cycle (Cycle 5) batch 6 fuel will j be loaded on the core periphery. In its third resident cycle <

(Cycle 6) the fuel will once again be loaded in the interior of the j core in a checkerboard arrangement, hence the term "in-out-in." i The fresh fuel (batch 6) will contain Burnable Poison Rod Asseitblies i (BPRA) to hold down local reactivity. The radial power distribution  !

will be tailored by employing three conce atrations of baron carbide.

Distributing the fresh fuel in the interior of the core, versus the periphery, reduces the neutron leakage. Therefore, for a <

fixed core enrichment, the cycle length will be increased. l Alternatively the cycle length may be increased by increasing the i average core enrichment. The designer has utilized both techniques I for Cycle 4 j

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The initial BPRA loading, '.he longer Cycle 4 design life and the variance in fuel management (in-out-in) make it difficult to compare the physics parameters between Cycle 3 and 4.

This is also believed to challenge the nuclear analytic capability to a greater extent than conventional three batch fuel management.

To insure that achieved power distributions in the core are within the confines assumed in the setpoint and safety analyses, periodic in-core power maps are to be taken (a current TS requirement) and power distributions compared with predicted values. The licensee has comitted to report the deviations as part of the plant's monthly operating report.

Power distribution control and reactivity control will be naintained by axial poser shaping rods (APSR), control rods and soluble boron concentration control. Figure 3-3 of Reference i shows the rod locaton, number of rods end function of each grouc. The Cycle 4 core will be operated as in Cycle 3 on a feed-and-bleed basis with the the APSRs inserted until near the end of cycle (E0C)

(approximately 20 EFPD).

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The projected Cycle a length is 335 EFPD with a predicted cycle burnup of 11,316 MW/MTV. The Cycle 4 nuclear parameters which include Doppler coefficients, xenon worth, critical boron con-centrations and effective delayed neutron fractions have been calculated using the approved P00-7 code (Reference 8). These physics parameters are presented in Table 5-1 of Reference 1 in conjunction with the Cycle 3 values for comparison. Relative to Cycle 3, the predicted critical boron concentrations are slightly decreased due to the use of BPRAs which tends to make the core

" blacker" to thermal neutrons thus offsetting the excess reactivity necessary for the longer cycle. The extended cycle will resu,lt in more fission products and hence a " blacker" core at end of Cycle 4 relative to Cycle 3. Small changes in the Doppler coefficient, moderator temperature coefficient, power defect and inverse boron worth are consistent with increased core blackness.

Shutdown margins have been calculated for beginning of cycle (B0C) and EOC (Table 5-2 of Reference 1). The calculated margin is 3.54% ak/k and 2.765 ok/k for BOC and E0C conditions respectively with the maximum worth rod stuck. These values are larger than the required 1% ak/k assumed in cooldown accident analyses by an adequate margin.

c 2.3 Thermal-Hydraulic Desian The Cycle 3 and Cycle 4 thermal-hydraulic design conditions for Rancho Seco are presented in Table 6-1 of Reference 1. The in-coming batch 6 fuel is hydraulically and geometrically similar to the fuel remaining from previous cycles with only a slight variation in active fuel length. The thermal-hydraulic method-ologies and models used to support Cycle 4 operation are described in References 6, 9, and 10. l 2.3.1 Core Bypass F. low In Cycle 3, the core bypass flow increased to 10.4% due to the removal of all orifice rod assemblies (ORAs). For Cycle 4. operation, 52 BPRAs will be inserted, leaving 56 vacant fuel assemblies and re-sulting in a decrease in calculated maxim;m core bypass flow to B.3%.

2.3.2 SPRA Retainers The inclusion of retainers to provide a positive holddown of BPRAs and neutron sources has introduced a small departure from nucleate boiling ratio (DNER) penalty as discussed in Reference 4 However, the increase in core flow as discussed above more than compensates for the decrease in DNBR due to the BPRA retainers.

2.3.3 Effect of Rod Bow on Thermal Desian The potential effects of fuel rod bow have been reviewed in Reference 11 in order to impose no DNBR penalty for fuel burnup .

to approximately 21,300 MWD /!!TU. That procedure was not accepted by the NRC staff and a modified procedure (Reference 11) that s imposes a non-zero DNBR penalty at 21,300 MWD /MTU, was agreed to by the NRC and B&W. The rod bow penalty applicable to Cycle 4 was calculated using the interim evaluation procedure as referenced above. According to the licensee, the calculated penalty using this procedure is less than 0.85. Utilizing the 1% DNB credit for.

the flow area reduction factor, no penalty is applied to the DNB calculations. Table 6-1 of Reference i lists the thermal-hydraulic parameters for Cycle 4 and the reference Cycle 3.

Based on the use of an approved rodel (Reference 12) and the 1 results of a bounding analysis , we concluoe that the licensee ]

has adequately tdcen fuel rod bowing into account f or the thermal design of Rancho Seco Cycle 4.

2.3.4 Control Rod Guide T'ube Wear By letter dated Noverber 23, 1979, the Coanission requested SMUD to provide cetailed inforcaticn on the wear characteristics of the control rods on the guide tuces in their fuel assemblies. , ,

In response, SMUD encaged B&W to perform confir:atory inspections l on selected control rod guioe tuees. The purpose of the inspec-  ;

tions was to provioe assurance tha the fuel could sustain the proposed Cycle 4 operation witnout experienci ng through-the-wall wear in the guice tubes, in accitien to provi cing generic information on B&W control roc gui:e tube wear. The results of the inspection program were previoed by SMU3's letter dated Feorua ry 15, 1950.

- The inspections mere :erf:roec by Eddy Currer. Test (ECT) tech-niques on dischargec fuel assemolies. Tne EFPD of cperation experienced by tne ir. spec;ec fuel ranged f rom 450 EFPD to 732 EFPD. The ECT easurements that were performed were calibrated with macninec stancarcs. Results of tne inspections '

incicated some loss cf wall thicktess due to wiar of the Zircaloy guide tuoes ey fretting action of :ne stainless steel -

clac control rocs in the :artec pcsition. No tnrougn-wall wear was observed in any cf :ne 127 guice tuoes exa-inec, anc only 64 guide tubes exniti ec neasuracle wear.

To provide adcitional assurance that guice tuce wear incicated by the test resuits woulc not affe:t tne structural integrity cf the fuel, B&W has reviewe: tne strengtn aspects of cegradation'.

The review consioerec uniform circumferential wear, one-sided wear, and two-sioea wear. Preliminary results indicate that the allowable wear in the limiting wear scenario would De in excess of 50% through-wall. Based on :ne guice tuce wear potential indicated by the ECT exandnations, and tne stress analysis results wnicn incluced wear oegra:atien, BiW conclucee :nat control rod guice tube wear coes not appear te De a signifi: ant pr blem in B&W fuel, and that Cycle ; coerati:n will not compromis: iM e structural integrity of tne fuel .

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l The staff agrees with the conclusions reached by the licensee and B&W to the extent that the ECT measurements appear to in-dicate that there is a sufficient margin between the actual wear observed or expected during Cycle 4 operation at Rancho Seco and design limitations.

3.0 Evaluation of Accidents and Transients General .

The licensee has stated that each accident analyzed in the Final Safety Analysis Report (FSAR) (Reference 9) has been examined with respect to changes in Cycle 4 parameters and has found to be bounded by the FSAR and/or the Fuel Densification Report (Reference 6) and/or subsequent cycle analyses.

The staff has concluded that the consequences of hypothesized events are no worse 'than those stated in the FSAR or previous submittals, That is, part 20 and part 100 cose rate limits will not be exceeded in the event of an Anticipated Operational Occurrence (A00). or accident respectively,nor wi,Il the margins to safety limits be significantly reduced.

Specific Analysis ,

The licensee has stated (Reference 1) that the generic B&W Emergency -

Core Cooling System (ECCS) analysis (Reference 13) is applicable to Rancho Seco, Cycle 4. Based on the minimal core changes for Cycle 4 the staff accepts this assertion.

The conclusion presented in the FSAR is that, in the event of a ,

Steam Line Break (SLB) accident, a small fraction of the 10 CFR 10'O dose rate would be reached. The supporting analysis assumed a 1% 66 safeguards allowance (shutdown margin). The predicted minimum shutdown margin during Cycle 4 is 2.76% _a6. On these bases the consequences of a hypothesized SLB are considered acceptable for .

Cycle 4 operation. -

The dropped rod accident analysis reported in the FSAR is based on an assumed dropped rod worth of 0.65t a V k, and a peak post dropped enthalpy rise, FaH, equal to the design value,1.71. The action

m of the Integrated Control System (ICS) is normally available to inhibit all rod-out motion and runback the steam generator load demand to 60 percent of rated load. Although ICS action is available to prevent or mitigate this accident, the accident analysis, according to the FSAR, was done without taking credit for ICS action. The licensee has predicted that the maxinum dropped rodworth is 0.2% ok/k. Post drop values of FaH were not provided by the licensee. However, the peak enthalpy rise would increase by less than 20% following the drop of a control rod worth only 0.2%a k/k, and this result is acceptable.

Following the rod drop and assuming no turbine runback, the core will return to rated pcwer. Since the core is typically operated witn an initial enthalpy rise approximately 15% (or greater) less than the design peak, and even if the core was initially at the design peak and the peak were to increase by 20% there would still exist margin to DNER limits (at 100% pcwer). The dropoed rod analysis is considered adequate for Cycle 4.

• The most limiting transient considered as a part of the original licensing process was the postulated loss of AC pcwer. The loss of all AC power would result in loss of reactor coolant pumps and loss of forced flow as well as loss of normal feedwater. This transient war analyzed, and its consequences are acceptable. The postulated loss of feedwater is gensidered less limiting than the loss of AC power assuming no other single or rultiple f ailures.

Therefore, loss of feedwater has not been reviewed as a part of the proposed license amendment.

The maximum ejected rod at hot full pcNer, Cycle 4, is predicted to be 0.39% AS. The FSAR analysis assumed a rod worth of 0.65". 46 The predicted Doppler coefficients during Cycle 4 are substantialsly less than the values used in the FSAR analyses. These are conservative changes relative to the FSAR analyses. The delayed neutron fraction (Seff) is predicted to be smaller than assumed in the FSAR. The effect of the smaller value of Seff is a slower decay of the neutron flux once the peak value is reached. This is a non-conservative change. The above cited conservatisms are substantially larger than this non-conservatism. FSAR calculations were run using a point kinetics design model assuming the design three dimension peak and compared to two dimensional space-time kinetics calculations. The design model was shown to be conserv-ative. The conclusion of this analysis presented in the FSAR is that there exist substantive margins for this accident to limit-ing enthalpy disposition values. On tnis basis the applicability of the FSAR analysis to Cycle 4 is acceptable.

1 The applicable loss of coolant accident (LOCA) analyses for Rancho Seco have been presented in Reference 13 which has been accepted by the staff for oeneric application to B&W 177-fuel assembly, lowered loop NSS plants.

The fuel densification report (Reference 10) describes the effect of densification on LOCA analyses and the use of the TAFY code ,

(Reference 7) to calculate fuel rod interval pressure and pellet volumetric average temperature.

We' concur with the licensee's assertion that the refueling of Rancho Seco for Cycle 4 will not result in kinetics parameters outside the bounds assumed for the FSAR analysis and that no change in the DNBR safety limit is required. The initial conditions for the transients in Cycle 4 are, according to the licensee, bounded i by the FSAR (Reference 9), the fuel densification report I (Reference 6), and/or subsequent cycle analyses. Additionally, the effects of fuel rod bowing and fuel densification on safety limits and on all transients and accidents', inclucing LOCA, have adequately been taken into account.

4.0 Emergency Core Cooling System An Exemption was granted on December 15,1978, to 10 CFR 50.46(a),

" Acceptance Criteria for Emergency Core Cooling Systems for Light i Water Nuclear Power Reactors." The Exemption provided for its own terrination upon completion of the modifications required by the Execption. The Exemption also found that the modifications, as - 4 pro;osed by the licensee, were acceptable. The licensee has installed tne modifications at Rancho Seco and has prepared acceptable operating procedures. Completion of testing of the modifications will take place before startup from the current outage as confirmed by the licensee in a letter dated March 21, 1980 (Reference 14). Therefore, we , conclude that the as-modified ECCS required by the Exemption of Decembhr 15, 1978, is acceptable.

5.0 Startuo Test Program A startup test program for the Rancho Seco Cycle 4 operation was submitted in the licensee's reload report. The licensee has comitted to provide the NRC with the results of the program within 90 days of completion of the tests.

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f.: Technical Specification Changes Proposed modifications to the Rancho Seco Technical Specifications (TS' are descriced below (Reference 1).

(1 ) The required quantity and concentration of Boric Acid I necessary to reach cold shutdown conditions have been I changed to be consistent with Cycle 4 requirements. i (2) Boron concentrations during fuel loading and refueling have l been set so as to remove restrictions on movement of control rod assemblies during refueling. Changes in the bases are indicated in order to be consistent with the Technical Specifications and Cycle 4 requirement.

(3) The following limits have been changed:

a. Core Prctection Safety Limits (FigJre 2.1-2)

b. Protective System Maxi num Allowable Setpoints (FigJ re 2.3-2)

c. Rod :ncex vs. Power ' e.el for Four-Punp Operation

(. Technical Specification FigJres 3. 5.2- 1, - 2, - 3) l d. Rod Incex vs. Power Level fcr Three-Pump Operation (Technical Specification FigJres 3.5.2 4, -5)

e. Rod !ncex vs. Power Level for Three-Pump Operation (Technical Specification FigJre 2.5.2-6)

f. A?SR Withdrawal vs. Dcwer Level (Technical Speci-ficatien Fi;J re 3.6.2-7, -3) 9 Core Im:alance vs. Ocwer Level Envelope (Technical Specifi:ation FigJres 3.5.3-9, -10, -11 )

The development of the revised safety limits, trip settings, co're imbalance limits, and control rod and APSR insertion limits was dor.e using the a: proved FLAP.E code (Reference 13) with the use of Cycle 4 peaking 'acters. The real allowable quadrant power tilt limit for Cycle 4 re .ains at the Cycle 3 limit of 4.92% maxiaum steady state.

7. 0 Enviror. rental C:nsideration We have deter-.ined that the amendment does net authorize a change in effluent types'cr total am:ur.:s ner ar. increase in power level and will not result in any significant environmental impact. Having made this ce:e .ination, we have further concluded that the amendment ir.volves an a::icn which is insignifican; fr:m the standpoint of erviron: ental i..:act and, pursuant to 10 CFR 151.5(d)(4), that an e vir:r er.tal i a:: state ent, or negative declaration and envir:n-

-me-tal '.,;a:: a:;raisal need n:: be prepared in conne: tion with the issuan:e of : tis amendmer.;,

8.0 Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a sicnificant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the heelth and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's -

regulations and the issuance of this amendment will not be inimical ,r tc the common defense and security or to the health and safety of the public. .

Dated: March 31, 1980

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References

1. Letter, J. J. Mattimoe (Sacramento Municipal Power Utility District) to R. W. Reid (NRC), dated October 3,1979, with attachment, Rancho Seco Nuclear Generating Station, Unit 1-l Cycle 4 Reload Report, BAW 1560, August 1979.

I latter,J. J. Mattimoe (SMJD) to R. W. Reid (NRC) dated 2.

February 19, 1980.

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3. Letter, J. J. Mattimoe (SMUD) to R. W. Reid (NRC) dated l February 27, 1980.

4 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.

5. Program to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse, BAW-10084, Rev.1, Babcock & Wilcox, Lynchburg, Virginia, Novemoer 1976.

6 Rancho Seco Unit 1 - Fuel Densification Report, BAW-1393, Babcock & Wilcox, Lynchburg, Virginia, June 1973.

7 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1973.

4 3

H. A. Hassan, et. al. , B&W Version of P0Q07 Coce, BAW 10117A, Babcock & Wilcox, Lynchburg, Virginia, January 1977.

9. Rancho Seco Nuclear Station, Unit 1 - Final Safety Analysis Report, Sacramento Municipal Utility District (Docket No.

50-312).

10 Rancho Seco Nuclear Generating Station, Unit 1 - Cycle 3 Reload Report, BAW-1499, Babcock & Wilcox, Lynchburg, Virginia, Septemoer 1973.

11. J. H. Taylor (B&W) to D. B. Vassallo (NRC), Letter, .l "Determinatien of the Fuel Rod Bow DNB Penalty," !l December 13, 1978. ]

12. J. H. Taylor (B&W) to S. A. Varga (NRC), " Determination of CHF Penalty at 55% Closure," June 22, 1979.

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13. ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Babcock & Wilcox, Lynchburg, Virginia, June 1975.

14. Letter, J. J. Mattimoe (SMUD) to R. W. Reid (NRC) dated fiarch 21, 1980.

15 FL AME - Three Dimensional Nodal Code for Calculating Reactivity and Power Distributions, BAW 10124A, August 1976.

, 16. Letter, R. W. Reid (NRC) to W. Cavanaugh, III, dated May 23, 1979.

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