ML20197D047
| ML20197D047 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/15/1978 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811210319 | |
| Download: ML20197D047 (10) | |
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$k u SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s street, Box 15830, sat.rartiento, California 95813; (916) 452-3211 n
't 347{ . -No vembe r 15, 1978 .
1
! , 3.d Director of Nuclear Reactor Regulation; ii fp
/
Attention: Mr. Robert W. Reid, ChiefW .i Operating Reactors, Branch a M 3 U. S. Nuclear Regulatory Commission '., - Washington, D. C. 20555 .: Docket No. 50-312! ,'" Proposed Angndment No. 60 . Rancho Seco Nire Station,UnitNog-Q,,.r Generating 7p. . *[* ['
Dear Mr. Reid:
Your letter of November 7,1978 requested additional information in support of Proposed Amendment No. 60 to Operating License DPR-54 for Rancho Seco Nuclear Generating Station, Unit No.1. This proposed amendment changes the Technical Specifications to permit operation with the fuel loading planned in Cycle 3. The requested information is attached to this letter. Sincerely yours, l
- , TL[ vt
'/J.Mb.
Mattimoe Assistant General Manager and Chief Engineer l Attachment l 1 l l 7611210n N I QE Ru' b]h s l
, AN ELLCTRIC sYSTE M S E R VIN G f.10 R L THAN C 0 0,0 0 0 th 1HE H l: ART Of calif 0PNIA
4
= ANSWERS T0~ QUESTIONS'0N PROPOSED' AMENDMENT'NO. 60 :
- 1. 1 question: Provide the maximum ejected ~ rod worths at the 10-3 .
1 power level and 100% power level for B0C and EOC conditions and state whether these are bounded by the values used in the FSAR. If these values- are not bounded by previous analysis provide the peak enthalpy for the hottest fuel rod.considering rod ejection at each of the above mentioned' power levels.' ; Response: The. zero power- ejected rod worths are provided in Table 5-1 of BAW-1499. The maximum full power ejected rod worth (a BOC. value)-_ is provided in Table 7 .1 of BAW-1499. The values listed for the maxi-mum ejected rods at the various conditions are well below FSAR values as stated in BAW-1499. For your information, the HFP,'E0C value is 0.27% Ap.
- 2. Question: Provide the post-ejection maximum peaking factors and compare these to what was used in the FSAR.
Respo_nse: os The methods and techniques for rod ejection accidents, which have been approved by the NRC, are described in the FSAR and do not' depend on ' post ejection peaking factors. The approved techniques have ber benchmarked against space-time-kinetics methods (See the FSAR).
- 3. _ Question:iState the reason -for withdrawal of the APSR during the '
20 days of Cycle 3. _ Resp,on se : The APSR's are being withdrawn during the last 20 day 1 cycle 3. to gain treactivi.ty which will increase cycle length, e . ovide
- a more' efficient utilization of. the fuel.
~
- 4. Question: Define the' axial stability index and describe how the numerical values presented on Page' 5-2 were obtained.
Response: Sufficient axial stability was demonstrated'by perturbing the xenon
- andfiodine con'centrations in several different ways and then calculating ~ the stability'index. . The stability index is the variable (b)'in the following equation: (which can be derived from Reference 1).
J l 2 OS.=:OSo+ebt[A] SINH +ACOS$] 2 whe re' OS = offset at time t i OSo= equilibrium offset T = period of oscillation In using this equation .to describe the perturbed offset behavior of the core, it is assumed that the offset is a sinusoidal function that decays exponen.ti ally. If b is positive, the oscillation- is divergent. A value of zero means an indefinitely sustained oscillation and a negative wlue , describes, a convergent one. ] The xenon and iodine were perturbed in .three ways:
'l) The APSR withdrawal procedure (Figure 1)
- 2) A 100-50-100 ' design transient.with return to full power ;
at peak xenon (Figure 2)-
- 3) A return .to full power with 50% power equilibrium xenon ]
and iodine concentrations.
. m . -_ _
The _ stability index was not calculated for. the first case,-but inspection of the offset behavior showed it to be convergent. The stability index for the second case was -0.051 hr-1, and -0.061 hr-1 for the third case. This amply demonstrates' tfe axial stability of the core. In addition, the maximum _imb'alance encountered after the return to full power was
-10%-and +10%. These are within the LOCA imbalance limits which are
-32.8 and +17.1 percent-imbalance.
The damping of the oscillation can be enhanced by insert Se control rods several percent as the imbalance goes through .its pos.:.i.e peak,
~
or withdrawing the rods as the imbalance goes through its negative peak. The effectiveness of this can be seen in figure 1. This figure shows the imbalance behavior with no rod movement after the return to full power from the APSR withdrawal (Case 2). The maximum positive imbalance encountered in case 2 is 4.4 percent less than in case 1. The subsequent negative peak is also less for case 2 by 1.1 percent.
- 5. Ques tio_n: State whether the axial stability index was calculated with !
NRC approved models. Response: The axial stability index was calculated as described in the i response to Question 4. The FLAME computer code (2,3) was used to i supply input to the calculation. References 1, 2, and 3 have been approved by the NRC.
- 6. Question: What is the maximum value of FAH predicted for this fuel cycle?
R_esjonse: The maximum theoretical value of FAh is 1.54 (no xenon peak) at BOC. This provides an 11% margin to the design value of 1.71 which l is more than adequate to cover calculational uncertainties. As indicated l in the response to Question 12 though, the actual maximum value of FAh at operating conditions (with xenon) is 1.51. Thus margin in excess of 13% exists. The max. predicted Fah decreases. to 1.35 at E0C for a 27% margin to 1.71.
- 7. Question: What is the basis for the selection of 1.71 for FAH?
Response: The 1.71 value W Fah was chosen for SMUD Cycle 3 because experience has she-
- nFAh of 1.7) is more realistic than the previously used 1.. 'l provides conservative margins to steady state and maneuverins limi ts . This value has been used for licensing Davis-Besse I Cycle 1, t, I Cycle 5, Oconee II Cycle 4, and Oconee III Cycle 4.
- 8. Question: In determining the flux / flow trip setpoint it is stated that an assumed one pump coast down was analyzed. Justify that this-is the worst event of this class when compared with a total loss of flow.
Response: The flux / flow trip setpoint is derived based on reaching the design minimum DNBR plus . applicable rod bow penalty during the one
; pump coastdown. The minimum DNBR used for the cycle 3 flux / flow set-point was 1.43 (B&W2) based on' the correlation limit plus 11.2% rod bow penal ty. Therefore, since the limiting.DNBR is used in the flux / flow
- analysu, it is; by definition the limiting flow coastdown. All other possible pump coastdown combinations will cause a reactor trip to be initiated by the pump' monitors. 0f these, the four pump coastdown is most limiting. The minimum DN3R during the four pump coastdown is 1
/
.a 4
- 1.660for(16% margin.to the design minimum.
- 9. Question: a. ilt:is noted that in the Technit al Specifications reactor -
coolant system flows previously stated in ibm /hr have 'been changed to gpm. _ Explain the need for'this change and verify that the new flows were calculated using the. reactor coolant density corresponding to the identified power levels,
- b. Explain why the title of' Figure 2.3-2 was changed from -
that:specified by the reactor vendor or restore it;to - its original wording. If the intent of this change in-
-wording is to allow maximum trip, setpoints_- above Curves --
1, 2 and 3 for. the. indicated pump configurations , provide analyses to' justify such higher setpoints. R_esponse: a. . Reactor coolant' system flows can be expressed equally well. in lbm/hr or gpm for various power / pump combinations. Flow rates in gpm are independent of rower level whereas lbm/hr calculations necessitate densi.ty corrections'. .Since ' reactor coolant pumps .are constant ' volume pumps gpm is more representative.of their operation. The values in gpm in the Tech Specs were calculated using the' reactor- coolant density corresponding' to the identified power levels. b._ The title of Figure 2.3-2 will be restored to its - original wording of " Protective System Maximum Allowable Setpoints, Reactor Power-Imbalance. " 10._ Question: Figure 2.1-2 ' Identify the specific credits taken in percent to allow the reactor power imbalance tent to change as' indicated. For example, what credits were taken to allow a 10% increase in power at
+50% and -50% axial imbalance .
Response: The credits taken in modifying the power imbalance tent-in Cycle 3 from that used in Cycle 2 were simply the differences in the three dimensional power distributions ' calculated for the two cycles.
- Table 1 below'shows the differences in FLAME total peaks (adjusted for radial-local, i.e. ,' hot pin to assembly average power) for the nominal fuel cycle depleti,on throu. gh- 100, EFPD.
c;
-TABLE l' Comparison of Cycles.2 and 3 Total
~
'- Peaking Factors r
TTime'(EFPDJ . Cycle 2 Cycle 3
- _ .t s0._ 7 1.93 1.85
~4' l.82' 'l.73 a25, ,
- 1. 81' 1.67-150: .
- 11. 8 0 / l'65.
, '100 1._76 1.63' L
[ I S. ,
- / w
4 s
- Further, comparison of similar extrene cases which led to the power imbalance tent limits .is shown .in Table 2.
TABLE 2 Comparison of Extreme Condition Cases for Cycles 2_ and 3 Cycle Negative Offset / Peak Positive Offset / Peak
'2 -35.3 2.75 16.9 2.17 3 -42.6 2.68 16.3 1.94 The imbalance limits derived from 1300 FLAME cases, including variations in burnup, xenon concentration, and control rod and APSR positions are given in Table.3.
TABLE 3 Comparison of Actual RPS i Imbalance Limits for Cycles 2. and 3 Power l L evel , Cycle 2 Cycle 3
%FP Neg. Imb. Pos. Imb. Neg. Imb. Pos. Imb.
. 112 -28.3 +16.1 -32.5 +35.8 100 -40.5 +44.1 -45.0 +60.0 80 -54.3 +70.1 -56.8 +83.2
- 11. Question: Quandrant Power Tilt - Justify the use of the values of ,
4.92%,11.07% and 20% for quadrant tilt on this reload. These values ' have been used in the Standard T/S as typical. Response: As stated in BAW-1499, page 8-1, the value of 4.92% actual quadrant power tilt as a steady state limit was accounted for by the use of a peaking factor of 1.0736 in the development of limits on control rod positions and imbalance. All values of the quadrant tilt ; limits proposed for use in cycle 3 have been reviewed and approved for i use in Davis Besse I, Cycle 1 and Crystal River III, Cycle 1. The same or larger values were approved for ANO-1, Cycle 3 and TMI-1, Cycle 4.
- 12. Question: Verify that the consequences of all the accidents and transients analyzed in the SAR are acceptable considering: (a)4.92% i tilt as an initial condition; and.(b) 20% tilt at 60% rated power as !
an initial condition. Response: (a) A 4.92% actual tilt would cause a_ maximum peaking increase of 7.36%,f as stated above. The maximum hot pin radial peak k
4 observed in Cycle :3 at operating conditions is 1.51 (assembly radial of 1.39 from Figure 5-1 of BAW-1499 times radial-local factor of 1.087). i The design radial peak used as .the initial condition for all accidents l and transients -(1.71) is 13% larger than the maximum hot pin for this- l cycl e. Thus, theiconsequences of all accidents are acceptable consider- 1 ing a steady state quadrant tilt limit of 4.92%. (b) When the reactor is operating at. initial power. levels less than rated power the initial steady state DNBR isL significantly larger than that at rated power operation. This increased initial , DNBR provides increased margin to the MDNBR limit for transients 1 initiated at less than rated power. . Four pump operation at rated - power represents the gieatest power to flow combination for initial l conditions:for DNBR evaluation following transients hnalyzed in the l SAR. Allowable power tilts of the magnitude presently considered for Cycle-3 have 'been found licensable for other 177FA B&W plants as enumerated in response to Question. ll. i
- 13. Question: Justify the changs to the core imbalance vs power level trip tents (Figures 8-11, 8-12 and 8-13 of BAW-1499) and explain why l the permissible operating region falls outside the maximum setpoint j values for reactor power imbalance shown in Figure 8-2. ;
i Response: The imbalance versur power level tents'shown in Figures l T391 5Fough 8-13 are derived cased on LOCA Kw/ft criteria and are " the actual limit values - not adjusted for measurement errors. The changes from Cycle 2 to Cycle 3 are the result of changes in the core i peaking characteristics, as discussed in the answer to Question 10. I Figure 8-2.shows the measurement error adjusted trip setpoints of the Reactor Protection System. In deriving these setpoints from the actual limits shown. in Table 3, many conservatisms are introduced. If, after the limits of Figures 8-11 through 8-13 are also error adjusted for alarm setpoints', they still fall outside of the trip setpoints, no problem arises. The result would simply be a trip signal prior to reaching a condition less severe than those for which the RPS provides protection. . "In' actual practice, the alarm setpoints would be moved I inside'of the' trip 'setpoints .to provide a warning to the operator that l a trip condition ~ Was being approached. l
REFERENCES:
l 1
- 1. Stability Margin' for Xenon Oscillations - Modal Analysis, BAW-10010, !
P_ art :1, Balkock and' Wilcox', August 1969. 1
- 2. ; FLAME ;Three-Dimensional Nodal Code for Calculating Reactivity !
and Power Distributions, BAW-10124A, Babcock &.Wilcox, August 1976.
- 3. Verification of Three-Dimensional FLAME Code, BAW-10125, Babcock &
Wilcox, August 1976..
- 4. ,0perational Parameters for B&W Rodded. Plants, BAW-10078, Babcock &
- Wilcox, September 1973.
i I 1
- Y
14; Question: ' Vour description of the ejected control rod reactivity worth test in BAW-1499 does not state that four symmetric control rods will be measured. As stated in BAW-1477."0conee 1 Cycle 4 Quadrant Flux. Tilt" page 12, this test "has proven to be an indicator of core
- symme try" . Please indicate if the measurement of ejected rod worth of four symmetric: location is part of your test program for the Cycle 3 core.
Response: Measurement of the ejected rod worth at four symmetric . locations is not part of the standard restart test program. Howeve r, the detailed zero power physics test procedure incorporates the requested measurement of symmetric ejected rod worths at four core locations.
- 15. Question: In Section 9.4 you describe actions to be taken if the -
Acceptance Criteria are not met. This description is too general for the.measumment of rod worth. The usual action to be taken if the sum of the worth of groups' 5, 6 and 7 differs from predicted by more than 110%, is. to measure group 4 by dilution and take additional measurements and make an evaluation of the. discrepancies. Indicate ^your commitment to thes'e corrective actions -if the acceptance criteria for measurement
~
of control bank reactivity worth are not met. Response: The : action to be ta' ken in the event the total measured worth of Groups 5-7, differed from the predicted value by'more than i10% is-to perform an; evaluation consisting of. one or more of the following
' items as appropriate to the situation:
- 1. Review of measurement data anf data analysis.
- 2. Verifjcation that the avilable shutdown margin based on the measured data satisfies the minimum shutdown margin require-ment.
- 3. Review of the. results of other physics test.
- 4. Review of calculations used to obtain the predicted value.
Evaluation of the impact of the discrepancy on safety of
- 5. -
operation and on Technical Specifications limits, if any.
- 6. Determination as to whether retest of one or more of the regulating groups would be required.
- 7. Determination as to whether measurement of one or more of the safety. groups would be required based on considerations of the. extent and nature of the discrepancy and of-item 5 above.
If it 'is determined' that measurement of one or more of the' safety groups would be required to resolve the discrepancy, such measurements will be performed.
1
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. j l
.16. Question: Indicate your commitment to submit a physics startup l test report within 45 days of completion of the tests.
Response: The results of the startup. physics program for Cycle 3 will be compiled and submitt'ed with one of the monthly. reports filed within 90 days after completi'on of the physics test program. Startup. j test data is availab'le for review at the station by NRC personnel prior to the fonnal submittal. i 17 Question: With regard to your request to modify T/S 3.3.3, the statement i that the proposed provision is like. the Standard Technical Specifications (STS) is not' sufficient justification for the change. The reason this is insufficient is because the STS are baseion a number of interrelated p rovi sions . Thus the adoption of a single provision without including j all of the related provisions may not be technically justified. Accord- ) ingly, if you desire to-make this change, you s1ould revise your sub-mittal as necessary to incorporate all STS operebility and surveillance requirements for engineered safety features in the Rancho Seco T/S, or propose a lesser revision supported by a detailed evaluation which demonstrates the safety of the new' provisions relative to the safety of the present provisions. Respons,e: The request to nodify . Technical Specifications Section 3.3.3 has been a subject of concern during Plant Review Committee discussions and NRC inspections. Both the Plant Review Committee and NRC inspectors ! agree that the proposed change in this' Technical Specifications will increase the safety of- operations. We believe that this provision of the Standard Technical Specifications contains a value judgment with which we concur; i.e., it is far safer to accept as a risk the very low probability that a safety features system, subject to a reasonable set of surveillance requirenents, wculd fail to start on demand rather than deliberately create a situation in which there is a certainty that both SFAS systens will be out'of i service' for a period of time. If after re-evaluation you do not concur with this judgnent, we will withdraw our request for a Technical Specification change. i i
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