ML20244D893

Safety Evaluation Supporting Amend 18 to License NPF-74

ML20244D893
Person / Time
Site:	Palo Verde
Issue date:	06/09/1989
From:	Office of Nuclear Reactor Regulation
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UNITED STATES y

{g NUCLEAR REGULATORY COMMISSION 5

l WASHINGTON, D. C. 20355

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1 SAFEiY EVALUATION BY.THE OFFICE OF NUCLEAR REACTOR REGULATION I

l RELATED TO AMENOMENT NO.13 T0. FACILITY OPERATING. LICENSE NO. NPF-74 i

k ARIZONA PUBLIC SERVICE COMPANY, ET AL.

1 PALO VERDE NUCLEAR GENERATING STATION, UNIT.NO. 3 DOCKET.NO. STN 50-530

1.0 INTRODUCTION

l By letter dated December 14, 1988 (Ref. 1), the Arizona Public Service Company

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(APS) on behalf of itself, the Salt River Project Agricultural Improvement and l

Power District, Southern California Edison Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority (licensees), requested several changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-74) for the Palo Verde Nuclear Generating Station, Unit 3 (PVNGS3), relating to Cycle 2 operation for PVNGS3.

In support of both the Technical Specification changes and Cycle 2 operation, the licensees submitted a Reload Analysis Report by letter dated December 27, 1988 (Ref, 2).

By letters dated April 26,1989 (Refs. 3 and 4), the licensees also provided corrections to the Reload Analysis Report necessitated by the revised end-of-cycle 1 termination burnup of 397 Effective Full Power Days, and infor-mation requested by the staff concerning the applicability of Reload Analysis Report references. The staff's evaluation of the reload analysis is presented in Sections 2.0 through 2.6 below. The evaluation of the specific changes to the Technical Specifications is presented in Section 3.0 below.

The reload will include 104 new Batch D assemblies, while 09 Batch A and 35 Batch B assemblies will be removed. All Batch C assemblies will remain for Cycle 2.

Cycle 2 will operate at the rated core power of 3,800 MWt. Throughout this submittal, Cycle I was used as a reference for Cycle 2.

The staff ha0; reviewed the submitted ir.f,rmation and the supporting documents n

regarding fueEdesign, nuclear design, i.hermal-hydraulic design, plant transients ard accident analysis. Our evaluation follaws.

2.0 EVALUAT!0ii 2.1 Fuel Mechanical. Design No changes in the fuel mechanical design basis have occurred in the fabrication of the Batch D fuel. Some design changes were made to improve fuel handling anc burnup capabilities of the poison rods. These changes involve:

(1) the upper end fitting hold down plate to improve handling, (b) a fuel assembly inspection envelope is changed to a square of 8.290 inches on the side for the 890620004t 890609 PDR ADOCK 05000530 P

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entire assembly length, (c) the poison rod assembly design was modified to replace the solid Zircaloy-4 spacers with hollow Zircaloy-4 tubes, and (d) replacement of the two piece lower end fitting center post with a one piece casting.

The staff has found Reference 3 acceptable where clad collapse analyses are not necessary for new Combustion Engineering manufactured fuel because of the absence of gaps between pellets.

We find the above changes to be minor improvements which do not affect the mechanical design basis and, thus, are acceptb le.

2.2 Thermal Design The Cycle 2 thermal performance evaluation was based on the performance of a composite fuel pin that envelopes the pin performance of all of the fuel present in Cycle 2, i.e., Batches B,(C, and D.Refs. 4-7) and a pover history enveloping The evaluation was performed using the NRC approved code FATES 3A the power and burnup levels representative of the peak pin at each burnup ini.erval from the beginning of cycle to the end of burnup (Ref.4). The peak pin burnup analyzed is in excess of that expected at the end of Cycle 2.

Based on this analysis, the internal pressure in the most limiting fuel rod will be 1,146.8 psia which is far below the reactor coolant pressure of 2,250 psia.

This satisfies the SRP requirements and is acceptable.

2.3 Nuclear Design 2.3.1 Fuel Management The Cycle 2 core will consist of 73 Batch B assemblies, 64 Batch C and 104 Batch D (new) assemblies. The Cycle 2 loading is low leakage, using previously burned-assemblies in the periphery. Thus, most of the Batch D assemblies are located throughout the core interior. The expected Cycle 2 lifetime is 410 effective full power days. The highest Batch D enrichment is 3.9 w/o U-235 which is lower than the 4.05 w/o U-235 for which the Palo Verde facilities have been 1

approved for fuel storage. Comparison of characteristic physics parameter for Cycle 2 and the reference cycle shows that the two cycles vary little from each other.

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2.3.2 Power

Distribution Calculated all-arb-out relative assembly power densities were provided for the beginning, raf Gle and end of cycle.

Relative assembly power densities for rodded configurations were also presented. The rodded configurations are those allowed by the power dependent insertion limit at full power. The nominal axial peaking factors are estimated to range from 1.14 to 1.11 at the beginning and end of Cycle 2, respectively. Augmentation factars have been eliminated from this cycle as discussed in Reference 8.

The w hodology for the physics t.nd power distribution calculations is based on RL,,,-DIT (with the MC module) which has been approved by the NRC (Refs. 9,10). These calculations based on approved methods are acceptable.

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2.3.3 Control. Requirements The most restrictive value of the shutdown margin occurs at the end of cycle under hot zero power conditions. The minimum shutdown margin required to control the reactivity transient resulting from a steam line break is 6.5%

delta-k/ k.

This shutdown margin is assured as discussed in paragraph 2.5.3.

In addition sufficient boration capability and control element assembly worth with a stock control element assembly exist to meet these shutdown requirements.

These results were derived with approved methods and incorporate conservative assumptions; therefore, the results are acceptable.

2.4 Thermal-Hydraulic Design Steady state thermal-hydraulic analyses for Cycle 2 were performed using the approved code TORC (Ref. 10), the Combustion Engineering CE-1 critical heat flux correlation (Ref. 11) and the CETOP code described in Reference 12. The methodologies described in References 10-12 with the statistical combination of uncertainties (Ref. 13) the core protection system, the core operating limit system and the DNBR value of 1.24 assure that at the 95/95 confidence / probability level the hot rod will not experience DNB. The 1.24 value includes all applicable penalties, such as the rod bow for burnups to 30,000 MWD /MTV, the.01 DNBR for the HID-1 grids and the penalties specified in the statistical combination of uncertainties (Ref. 14-16). The rod bow value used in the analysis is 1.75% DNBR, for burnups up to 30,000 MWD /MTV. For burnups higher than 30,000 MWD /MTU sufficient margin exists to offset the rod bow penalty due to lower radial power peaks in these higher burnup assemblies and rods, hence, the rod bow penalty is adequate for all anticipated burnups.

We conclude that the thermal-hydraulic design analyses were performed using approved codes and accounted for all applicable penalties, e d therefore, are acceptable.

l 2.5 Safety Analyses.(Non-LOCA)

The design basis events considered in this safety analysis are classified in two groups: The anticipated operational occurrences (moderate frequency and infrequent events) and the limiting fault events i.e., postulated accidents.

All events were; evaluated with respect to four criteria: fuel performance (centerline melt), reactor coolant system pressure, loss of shutdown margin and offsite dose. All events were reevaluated to assure that they meet their respective criteria for Cycle 2.

The limiting events for each criterion and those not bounded by the Cycle 1 values were reanalyzed. The analytical l

methodology for the reanalyses are the same as for Palo Verde Unit 3 Cycle 1 l

1.e., the reference cycle as described in the FSAR, the CESSAR and Ref.17.

All of the methodologies used have been reviewed and approved by the NRC. The l

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4-following list includes the code, the purpose for which was used in the analyses and the reference:

Code Purpose Ref.

CESEC-III Plant response to non-LOCA events 17 CETOP-D Hot channel and DNBR 12 TORC Pin DNBR and RCP shaft seizure 10,18 CENPD-183 1.oss-of-flow methodology analysis 19 HERMITE Core simulation for space-time kinetics 20 The input parameters for the analyses were comparable to those for the reference cycle. Whenever the core protection system trip was evoked in the sequence the instrument channel response times assumed were conservative relative to the applicable Technical Specifications.

All of the events evaluated are bounded by the reference cycle except:

control element assembly misoperation, asymmetric steam generator events and steam line break, Och of these limiting events analysis is discussed below.

2.5.1 Control Element Assembly Misoperation A control element assembly (CEA) misoperation is defined as the inadvertent release of a single CEA (or a CEA subgroup) causing it to drop into the core.

The single full length and part length control element assemblies were reanalyzed to determine the initial thermal margin that must be maintained by the limiting conditions of operation such that the DNBR and the full centerline melt limit will not be violated. Because the control assembly downward position penalty factors have been eliminated, a 4-fingered control assembly drop will not generate a trip, therefore, sufficient margin must be maintained by the limiting conditions of operation. However, for the 12-fingered control assemblies the core protection calculator will provide a trip.

The method used to analyze the single control assembly drop event is

' described in Reference 21.

The single full length CEA was analyzed because this event requires the maximum initial margin to be maintained by the limiting conditions of operation. Thelinitial conditions were selected conservatively, for example:

the turbine load is not reduced, causing a power mismatch, the cycle most negative moderator and fuel temperature coefficients were assumed, the charging pumps and pressurizer heaters are assumed inoperable (to maximize pressure drop) and all other systems are assumed to be on manual mode having no effect on the transient. The event is initiated by dropping a full length CEA over 1.0 sec. The largest power peaking was obtained by examining the configurations allowed by the power dependent insertion limit, which resulted in a peaking factor increase of 8.5% and a minimum DNBR greater than 1.24 at 15 minutes into the event.

However, before this time i.e., at 10 minutes the operator will take action to reduce power if the CEA has not been realigned, according to Fig. 3.1-2A of the Technical Specification. A maximum allowable

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initial heat generation rate of 18.0 Kw/ft could exist as an initial condition without exceeding the acceptable steady state fuel centerline melt limit of 21.0 Kw/ft.- The heat rate for this transient is based on the more limiting LOCA initial heat rate of.13.5 Kw/ft.

Considering the above we conclude that the control assembly misoperation analysis meets the requirements of the SRP Section 15.4.3 and is acceptable.

2.5.2 Assymmetric. Steam Generator. Events Of the four events which could affect a steam generator it has been determined that the loss of load to a single steam generator is the most limiting event.

Such total loss of load could result if both main steam isolation valves were inadvertently closed. This sequence'of events causes an initial increase of temperature and pressure and decrease of the water. level of the affected steam generator. Pressure will increase, which might cause the secondary safety valves to open.

In addition the steam generator low level trip may be activated. On the primary side the increased water temperature will result in a temperature tilt across the core with power increases on the cold side which could potentially approach the DNBR and the fuel melt limits. The core protection system, has a high cold leg differential temperature trip which is the' primary means of protection.

This eve.it is analyzed using the CESEC code for the NSSS response (Ref.17).

The resulting case parameters form the input to the HERMITE code, to model the effects of the space-time radial power tilt (Ref. 20). Finally the thermal margin changes are evaluated using the CETOP code (Ref. 12).

The results indicated that the minimum DNBR is greater than 1.24, based upon a high differential cold leg trip being generated at 6.0 second into the event.

.We find the methods and the results acceptable.

2.5.3 Steam Line Break A steam line break can take place inside or outside containment.

Inside containment may cause environmental degradation to sensor inputs and/or the core protection _ system. However, the variable overpower trip, which includes inputfromthefrosistancetemperaturedetectorsandtheexcoredetectors,can be assumed functional under all conditions.

y The case which was found to be less conservative with respect to the reference j

cycle is the steam line break at zero power, because the cooldown reactivity insertion curve is more adverse than the reference cycle curve.

In addition, a sweepout volume of 119 ft3 before safety injection reaches the RCS was assumed for the post-trip return to power case as opposed to the 34.7 fta accounted for in the reference cycle. The effect of this reactivity was accommodated in Cycle 2 by increasing the shutdown margin required by the i

Technical Specifications at zero power from 6% delta-k/k to 6.5% delta-k/k.

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Considering the above results we find the steam line break analysis acceptable.

-2.6 ECCS Analyses An.ECCS analysis was performed for the limiting break size LOCA for Cycle 2 to demonstrate compliance with the requirements of 10 CFR 50.46. The methodology is the same as for the Cycle 1 analysis (Ref. 22). The analysis justifies a 13.5 Kw/f t peak linear heat generation rate. For Cycle 2, since there have been no significant changes in hardware characteristics, only clad temperatures and oxidation are required in this reevaluation. The code STRIKIN-II was used for this purpose (Ref. 23). The performance data were generated with the FATES-3A fuel evaluation code (Refs. 5 & 6).

It was demonstrated that the double ended guillotine break with a discharge coefficient of 1.0 is the limiting size. Similarly the limiting burnup, i.e., with the highest fuel stored energy, was found to be at 1,000 MWD /MTU. The ECCS analysis methods discussed above have been previously approved and are acceptable.

2.6.1 Large LOCA Analysis The input data compared to the reference cycle were conservative. The results for the limiting double ended guillotine break showed a peak clad temperature of 1,957'F, peak clad oxidation, 5.8% and total core-wide oxidation less ~than -

.80%.

All these values are within the required 10 CFR 50.46 limits of 2,200'F, 17.0% and 1.0% respectively. Therefore, we find the large LOCA analysis results to be acceptable.

2.6.2 Small Break.LOCA. Analysis

- Review of the Cycle 2 fuel and core data confirmed that the small break LOCA dnalysis results are bounded by the corresponding results of the reference cycle.

3.0 TECH M AL SPECIFICATION CHANGES This section provides a summary of the proposed amendments to the Palo Verde Unit 3 Technical Specifications for the Cycle 2 operation. A brief description, justification and acceptability for each Technical Specification (TS) change is provided in the;following.

TS 3.1.1.2:

The proposed change raises the required shutdown margin for hot zero power conditions from 6.0% delta-k/k to 6.5% delta-k/k to accommodate the requirements for the steam line break. This change is necessary to satisfy regulatory requirements and thus, is acceptable.

TS Fig 3.1-1: The main change is a broadening of the moderator temperature coefficient (MTC) range. Another is the change of the abscissa from average moderator temperature to core power. The increased MTC range is necessary from the increased enrichment of the new fuel. The maximum positive MTC at low power is 5.0 pcm/"F, (1 pcm=10~5 delta k/k). This value satisfies accident

analyses requirements, is within the limits of the ATWS requirements (Ref.16) and satisfies GDC 11, therefore, this change is necessary and acceptable. The change of the abscissa is a matter of convenience, is not substantive and is acceptable.

TS 3.2.8: The proposed change affects the operational pressure band for tha pressurizer pressure, from 1,815-2,370 psia to 2,025-2,300 psia. The reason is to support the core protection calculator improvement program. The proposed change narrows the permissible band and is, therefore, conservative ano acceptable.

TS.3.1.3.1, 3.1.3.2, 3.10.2 and.3.10.4: The proposed change separates the part length control element assemblies (PLCEA) from TS 3.1.3.1 and 3.1.3.2 to a separate TS 3.1.3.7.

In addition the special test exception for the PLCEA are modified to reference the new TS 3.1.3.7.

The proposed change constitutes an improvement over the existing TSs, because it is an explicit reference to the PLCEAs, allowing to differentiate their operational requirements. Therefore, these changes are acceptable.

TS 3.3.1 Table 3.3-2:

The proposed amendment changes the response time of the DNBR low reactor coolant pump shaft speed trip in the Technical Specification Table 3.3-2.

The change from 0.70 sec to 0.30 sec response time was necessitated by the redefinition of the trip condition. The shaft speed trip redefinition is proposed to eliminate unnecessary spurious trips of the system. We find this change an improvement and, therefore, acceptable.

TS 3/4.1.3.5 and 3/4.1.3.6.and Fig. 3.1-3.and 3.1-4:

The proposed amendment constitutes a revision of the existing 15 to address control element assembly (CEA) insertion limits with one or two CEAs out of service. This revision was necessitated by the change in the CEA worth in the Cycle 2 core physics. The revisions are necessary to maintain shutdown margin, therefore, are acceptable.

TS Tables 3.3-2.and 3.3 2a:

The proposed amendment eliminates core protection calculator penalties which compensate for the resistuca temperature detector response times greater than 8 sec. The Cycle 2 safety analyses assume a maximum resistance temperature detector response time of 8 sec and do not include calculator penalty factors for response times greater than 8 sec. The removal of the penalty factors is acceptable because these factors are nof.:used in the Cycle 2 safety analyses.

TS 2.1.1.1,. Table.2.2-1, Bases.2.1.1.and.2.2.1*: The proposed amendment changes references of the calculatea departure from nucleate boiling ratio i

(DNBR) from 1.231 to 1.24.

This amendment also deletes references to the calculation of additional rod bow penalties if the penalty incorporated into the DNBR limit is not sufficient for any part of the cycle.

The methodology and the results discussed in section 2.5.2 yielded a DNBR limit of 1.24.

This value continues to ensure that power operation limits A This specification change should include DNBR change to 1.24 in the Bases of 3/4.4.1 which was not incl.ded in the licensee submittal request.

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'* iy calculations are conservative. This is acceptable. As for the rod bow penalty, it is now more advantageous to use a large enough DNBR penalty to provide protection throughout the cycle, rather than to apply incremental penalty f actors., A rod bow penalty factor of 1.75% DNBR provides sufficient margin for burnups up to 30,000 MWD /MTV.

For greater burnups, there is sufficient margin from other factors to offset any small increase in the rod bow penalty.

TS 3.2.5:.Th's proposed amendment changes the reactor coolant system (RCS) 6 total flgw f rom greater or equal to 164.0x10 lbm/Sr to greater or equal to 155.8x10 lbm/hr. This change is consistent with the value of flow used in the plant safety analyses and is therefore acceptaole. The wording of the specification has also been modified to clarify the need to account for instrument error in the comparison of measured flow to the safety analysis value. This clarification is an editorial enhancement and is, therefore, acceptable.

TS 3/4.2.1:

The proposed amendment changes the linear heat rate (LHR) limit from 14.0 Kw/ft to 13.5 Kw/ft. This change was the result of the safety reanalysis for Cycle 2 in order to ensure that peak clad temperature does not exceed the 10 CFR 50.46 limits.

In addition the amendment delineates how the LHR is to be monitored. This TS change is Lased on acceptable safety analysis and, therefore, is acceptable.

TS 3/4.2.4,. Table-3.3 1,.Baus 3/4.2.4.and.3/4.1.3:

There are a number of proposed amendments to ensure reactor operation within approved safety limits, increase operator reliability and increase clarity. The proposed changes in each.TS are discussed below.

Section 3.2.4 is replaced by a new format which addresses the specific conditions for monitoring)DNBR with or without the core operating limit supervisory system (COLSS and/or the control element assembly clusters.

These cover the following four cases:

(a)COLSSisoperableandeither orbothCEACsareoperable(b)COLSSoperablebutneitherCEACis operable (c) COLSS inoperable and one or both CEACs are operable and (d)

COLSS out of service and neither CEAC is operable. The new format states the action statements for all four of the above cases and replaces Figures 3.2-1 and:3.2-2 with new Figures 3.2-1, 3.2-2 and 3.2-2a.

Sectionb-1-Thissectionimplementsthefollowingchangesintroducedin 3/4.2.4 above i.e., (a) removes references to reactor operation with or without COLSS operable and both CEACs inoperable and (b) deletes the graph of DNBR operating limit margin (fig. 3.3-1) based on COLSS for both CEACs inoperable.

(This is a result of this information having been incorporated into TS 3/4.2.4)

Section 3/4.1.3 Bases, changes to reflect the changes in 3/4.2.4 above.

Section 3.2.4 Bases, change to reflect changes in 3.2.4.

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L The above changes ensure operat~

of Cycle 2 within the approved safety analyses and increase the speci

'tions' wording clarity. Therefore, they are acceptable.

l Figure 3.2-4a. This amendment modifies the azimuthal tilt operating limits with the core operating limit supervisory system in operation, to avoid i

lengthy delays in increasing power. When the core operating limit supervisory system is in operation, reactor operation within the analysis limits is assured, therefore, the proposed amendment is acceptable.

TS 3.3.2, Table 3.3-4:

The proposed modification removes the " greater than" sign from the trip value of the refueling water storage tank actuation signal in Table 3.3-4.

This removes an ambiguity concerning the level setpoint and is acceptable.

TSs 3/4.3.1, 3/4.3.2 and 2.2.1, Bases: The proposed changes constitute administrative changes to the bases of 3.3.1, 3.3.2 and 2.2.1 to ensure clarity and conciseness, to include updated references and to remove Cycle 1 information no longer applicable to Cycle 2.

These cFanges are, therefore, acceptable.

4.0 STARTUP TESTING The licensee presented a description of the planned startup testing, which includes:

low power physics, ascension to power and procedures if acceptance criteria are not met. The objective of the testing is to verify that the core performance is consistent with the design and safety analyses. The program lements the conforms to the requirements of the ANSI /ANS-19.6.1, 1985 and supp(Refs. 24 &

normal surveillance requirements of the Technical Spv ifications 25). The lower power physics tests include:

initia, criticality, critical boron concentration, temperature reactivity coefficient, control element assembly reactivity worth and inverse boron worth. The power ascension testing includes:

flux symmetry verification, core power distribution, shape

-annealing matrix, boundary point power correlation coefficient, radial peaking factors, control element assembly shadowing factor, reactivity coefficient at power and critical boron concentration. These tests will provide reasonable assurance that the core has been loaded in accordance with the safety analysis assumptions. They are therefore acceptable.

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Should any of*tlE startup tests reveal any unreviewed safety issues the NRC will be notified.

5.0

SUMMARY

AND CONCLUSIONS

.The Reactor Systems Branch reviewed the submitted information in support of the Palo Verde Unit 3 Cycle 2 operation. The review covered fuels, physics, thermal hydraulics, accident and transient analyses, technical specification revisions and startup test procedures.

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• Based on the evaluations presented in the preceding sections we find the proposed reload acceptable.

6.0 CONTACT WITH STATE OFFICIAL The Arizona Radiation Regulau.ry Agency was advised of the proposed determi-nation of no significant hazards consideration with regard to these changes.

No comments were received.

7.0 El4VIRONMENTAL CONSIDERATIONS This anendment involves a change in the installation or use of facility com-ponents located within the restricted area as defined in 10 CFR 20 relating to a reactor refueling. The staff has determined that this amendment involves no significant increase in the amount, and no significant change in the type, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational rrdiation exposure. The Commission has previously issued proposed findings that the amendment involves no significant hazard consideration, and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessnent needs to be prepared in connection with the issuance of this amendment.

8.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that (1) be endangered by operation in the proposed manner, (2) y of the public will not there is reasonable assurance that the health and safetsuch activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. We therefore, conclude that the proposed changes are acceptable.

Principal Contributor:

L. Lois Dated:

June 9, 1989

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9.0 REFERENCES

1.

Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station Unit 3, Proposed Reload Technical Specification Changes," dated December 14, 1988.

2.

Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station Unit 3, Submittal of the Reload Analysis Report," dated December 27, 1988.

3.

Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC " Generic Applicability of EPRI NP-3966-CCM, Volume 5"(161-01867-JRP/DBK} dated April 26, 1989.

4 Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "

Operating History of tne Reference Cycle" (161-01866-DBK/JRP) dated April 26, 1989.

5.

CENPD-139-P-A, "C-E Fuel Evaluation Modol," Combustion Engineering, dated July 1974 6.

CEN-161(B)-P, " Improvements in the Fuel Evaluation Model," Combustion Engineering, dated July 1981.

7.

Letter from R. A. Clark (NRC) to A.E. Lundvall, Jr. (BG&E), " Safety EvaluationofCEN-161(FATES 3),"datedMarch 31, 1983.

8.

CENPD-153P, Rev.1-P-A, " INCA /CECOR Power Peaking Uncertainty,"

Combustion Engineering, dated May 1980.

9.

CENPD-266-PA, "The ROCS and b'T Computer Codes for Nuclear Design,"

Combustion Engineering, dated April 1983.

10. CENPD-161-PA, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," Combustion Engineering, dated Aprii 1986.

11. CENPD-162-A, " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard $ pacer Grids, Part 1, Uniform Axial Power Distribution,"

Combustion. Engineering, dated September 1976.

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12. CEN-160-5,' Rev.1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Unit 2 and 3," Combustion Engineering, dated September 1981.

13.

CEN-356-V-PA, Rev. 1-PA, " Modified Statistical Combination of Uncertainties," Combustion Engineering, dated May 1988.

14. CENPD-225-PA, " Fuel and Poison Rod Bowing," Combustion Engineering, dated June 1983.

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15. Letter from A.E. Scherer Combustion Engineering to D.G. Eisenhut NRC (Enclosure 1), " Statistical Combination of System Parameter Uncertainties

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in Thermal Margin Analyses for System 80," dated May 14, 1982.

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16.

CESSAR SSER 2 Section 4.4.6, " Statistical Combination of Uncertainties,"

Combust %n Engineering.

17. CESEC, " Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," Combustion Engineering enclosure 1-P to LD-82-001, dated January 6,1982.

18.

CENPD-206-P, " TORC Code Verification and Simplified Modeling Methods,"

Combustion Engineering, dated January 1977.

19 CENPD-183, " Loss of Flow, CE Method for Loss-of-Flow Analysis,"

Combustion Engineering, dated July 1975.

20.

CENPD-188-A, "HERMITE, Space Time Kinetics," Combustion Engineering, dated July 1975.

21.

CENPD-199-PA, Rev. IP, "CE Setpoint Methodology," Combustion Engineering, dated January 1986.

22, CENPD-132-P, "Calcu?ative Methods for the CE Large Break LOCA Evaluation Model," Combustion Engineering, dated August 1974. Also Supplements 1 and 2 dated December 1974 and July 1975 respectively.

23. CENPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," Combustion Enginee. ring, dated April 1974. 7,150 Supplements 2P and 4P dated February 1975 atd August 1975 respectively.

24. ANSI /ANS-19.6.1-1985, " Reload Startup Physics Tests for Pressurized Water Reactors."

25. CEN-319, " Control Rod Group Exchange Technique," Combustion Engineering, dated Novembe.* 1985.

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