ML19319E074
| ML19319E074 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/28/1973 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| References | |
| NUDOCS 8003310585 | |
| Download: ML19319E074 (50) | |
Text
-
- - ~
/
s u _n, m
.$, T M
n.rr-p-
1_.
s; ; J :.1~ C' ' ~..
c s.
^fgGM W3 SUPPLEMENT NO. 1 TO THE SAFETY EVALUATION BY THE DIRECTORATE OF LICENSING U. S. ATOMIC ENERGY COMMISSION IN THE MATTER OF SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT NO. 1 DOCKET NO. 50-312 NOV 2 8 573 RETURN TO REGULATORY CENTRALFITRS ROOM 016 8 0 03 310 g85 2 r-
.a w.
4
' s, TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
1-1 2.0 FUEL DEtlSIFICATION 2-1 2.1 General 2-1 2.1.1
Background
2-1 2.1.2 Scope of Review 2-2 2.2 Mechanical Integrity of Cladding 2-3 2.3 Effects of Densification on Steady State and 2-4 Transient Operation 2.3.1 General 2-4 2.3.2 Fuel Rod Thermal Analysis 2-6 2.3.3 Steady State and Loss-of-Flow Transient 2-7 2.3.4 Other Transients 2-9 2.3.5 Summary 2-10 2.4 Accident Analyses 2-10 2.4.1 General 2-10 2.4.2 Locked Rotor Accident 2-13 2.4.3 LOCA Analysis 2-14 2.4.3.1 Core Axial Flux Profile 2-16 2.4.3.2 Moderator Temperature Coefficient 2-17 2.4.4 Rod Ejection Accident 2-17 2.5 Summary and Conclusions 2-19 2.6 References 2-22 i
y r
~.
y
TABLE OF CONTENTS (Continued)
PAGE 3.0 LARGE, SMALL AND CORE FLOODING TANK LINE BREAKS - LOCA's 3-1 3.1 Large Break Analysis 3-1 3.2 Small Break Analysis 3-3 3.3 Core Flooding Tank (CFT) Line Break Analysis 3-4 3.4 Conclusions 3-5 3.5 References 3-6 4.0 OTHER MATTERS 4-1 4.1 Diesel Generator Schematic Review 4-1 3
4.2 Emergency Feedwater System 4-1 5.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 5-1 5.1 Operation at 2772 MWt 5-1 5.2 Fuel Loading Procedures 5-2 5.3 Tritium Management 5-2 5.4 Positive Moderator Temperature Coefficient 5-3 5.5 Pump Overspeed 5-3 5.6 Common Mode Failure and Anticipated Transients Without Scram 5-3 5.7 Course-of-Accident Instrumentation 5-4 5.8 Applicant's Management Safety Review Committee 5-5 5.9 Control of Power Peaking Factors and Linear Heat Rate 5-6 5.10 Changes in AEC ECCS Acceptance Criteria 5-6 5.11 References 5-7
6.0 CONCLUSION
S 6-1
APPENDICES Appendix A Chronology....................... A-1 Appendix B Report of Advisory Committee on Reactor Safeguards... B-1 111
j 1-1 i
1.0 INTRODUCTION
The Sacramento Municipal Utility District (hereinafter referred i
to as applicant) by application dated November 20, 1967, as subsequently amended, requested a license to construct and operate a pressurized water reactor, identified as the Rancho Seco Nuclear Generating Station Unit No. 1 at a site located about 25 miles southeast of Sacramento, California. The Atomic Energy Comission Regulatory staff reported the results of its review prior to construction in a Safety Evaluation Report dated August 12,.1968. Following a public hearing before an Atomic Safety and Licensing Board in Sacramento, California, on September 17, 1968, the Comission iss,ued Provisional Construction Permit CPPR-56 on October 11, 1968.
April 29, 1971, the applicant filed, as Amendment No. 5, the Final Safety Analysis Report (FSAR) required by Part 50.34(b) of Chapter 10 of the Code of Federal Regulations as a prerequisite to obtaining an operating license for the facility.
The Comission reported the results of its operating license review in a Safety Evaluation Report dated June 8, 1973. This supple-ment to that Safety Evaluation Report documents the conclusions of the Regulatory staff review of(l) matters identified in the June 8,1973 Safety Evaluation Report as items still under review, and (2) matters covered in the report of the Advisory Comittee on Reactor Safeguards (ACRS).
This report also contains an updated chronology as Appendix A and the report of the ACRS as Appendix 8.
M m.
2-1 2.0 FUEL DENSIFICATION 2.1 General 2.1.1
Background
On November 14, 1972, the Regulatory staff issued a report entitled, " Technical Report on Densification of Light Water Reactor Fuels"II)* which resulted from the staff's consideration of the Ginna fuel densification phenomenon. Based upon the findings in this report the staff requested on November 20, 1972 that the applicant provide analyses and relevant bases, in accordance with thedensificationreport,III that would determine the effects of fuel densification on normal operation, transients and accidents for the Rancho Seco Unit I facility. On January 16, 1973, the Duke Power company filed a response to the request for Oconee Unit 1 as a lead plant for this evaluation.(2,3) On March 14, 1973, the staff requested additional information. Duke Power Company filed a response j
to this request on April 13,1973.(4,5) On June 8, 1973, the applicant filed a response to the request.(6)
The staff's technical review of fuel densification as it applies to Rancho Seco Unit 1, and its technical evaluation of the applicant's safety analysis of steady state operation, operating transients and postulated accidents taking into account the effects of densification are presented in this supplement. This evaluation relies upon the July 6,1973 Regulatory staff report entitled, " Technical Report on
- Numbers in ( ) refer to references listed in Section 2 6
2-2 Fuel Densification of Babcock & Wilcox Reactor Fuel"(7) which concluded that B&W's fuel densification models, as modified, are in compliance with the staff's initial densification report.II)
The staff has concluded that the operation of Rancho Seco Unit 1 for the first cycle at power levels up to 100 percent of full power, in accordance with the Technical Specifications, will not present an undue risk to the healtt. and safety of the public.
2.1.2 Scope of Review The essential elements that must be considered in evaluating the effects of fuel densification have been set forth in the staff's initial densification report.II) Since the performance of the facility in steady state operation and during various postulated transients and accidents had been established previously as reported in the Final Safety Analysis Report (FSAR) without the assumption of fuel densification, it was only necessary to evaluate those changes in the analyses and in the results that are attributed to fuel densification. The effects of fuel densification on the steady state operation and on the course of postulated plant transients and accidents were evaluated by the applicant l
and reviewed by the staff.
i The staff reviewed the effects of fuel densification for Rancho Seco Unit I using the staff's guidelines, the technical r
~
2-3 evaluation of the applicant's safety analysis of steady state operation, operating transients and postulated accidents and the generic evaluationI7) of B&W methods for assessing fuel densifica-i tion and its effects.
In the evaluation, the applicant appropriately considered the staff guidelines, including the effects of instantaneous and anisotropic densification (initial density minus 2a, and final density 96.5% TD), the assumption of no clad creepdown as a function of core life, and the assumption of an axial gap leading to a power spike. The staff reviewedII) the effects of fuel manufacturing and reactor operating parameters on the fuel densification mechanism.
Thestaffreviewed(7)B&W'sassumptions, methods,andcomputercodes used in evaluating the fuel densification effects. The mechanical integrity of the fuel cladding and the thermal perfonnance of the fuel were considered in the analyses of steady state operation, operating transients, and postulated accidents as discussed in the following sections.
2.2 Hechanical Integrity of Cladding Clad creepdown during the core life is not considered by the applicant in the calculation of gap conductance. This is a conser-vative assumption since the reduced gap size due to clad creepdown would result in a higher gap conductance and thus in a lower stored energy in the fuel. The staff reviewed the B&W method for calculating the clad collapse time, which is the time required for an unsupported 6
+,,
4
2-4 cladding tube to flatten into the axial gap volume caused by fuel densification. On the basis of independent staff calculations and from experience of similar fuel in other reactors, the staff concurred with the applicant that clad collapse is not expected for the Rancho Seco Unit 1 fuel during the first cycle of 12,000 effective full power hours (EFPH). However, the staff concluded that tne evaluation model for collapse time calculations contains several deficiencies in its application to Rancho Seco Unit i. The staff informed the applicant (8) that an acceptable model for collapse time calculations will be necessary for its evaluation of subsequent fuel cycles.
2.3 Effects of Densification on Steady State and Transient Operation 2.3.1 General Fuel densification can affect steady state operation because axial gaps in the fuel column result in local neutron flux spikes and an overall increased linear heat rate. An additional effect occurs in the transient analyses since, due to a lower gap conductance, the fuel has a higher initial stored energy and a slower heat release rate during the transient.
On the basis of evaluations of the effects of fuel densifica-l tion, the Rancho Seco Unit I reactor will be operated with more restrictive limits on control rod patterns and position than originally proposed, and with a reduced maximum linear heat generation rate.
^
l 2-5 l
The limits are based on consideration of the effects of local peaking caused by gaps in the fuel pellet stack and changes in gross peaking factors, primarily axial, which can be achieved by more restrictive operation of control rods.
The effects of densification on power density distributions have been calculated using models in conformance with those discussed in Section 4 of the staff densification report.II) The primary cal-culations used the models and numerical data of the Westinghouse power spike model as described in Appendix E of that report, except that initial nominal density used was [ ]* percent of theoretical density and the probability of gap size was changed to confonn to that recommended by the staff.(I)
The calculations by the applicant take into account the peaking due to a given gap, the probability distributien of the peaks due to the distribution of gaps, and the convolution of the peaking probability with the design radial power distribution. The calcula-tions result in a power spike factor that varies almost linearly with core height and reaches a maximum value of 1.10 at the top of the core. The overall calculation falls within the range examined (9,10) by our consultant, Brookhaven National Laboratory, in conjunction with reviews of other models.
A normalized shape for the power spike factor is derived from power spikes caused by different gap sizes at various axial locations.
The nonnalized shape is then used in conjunction with various axial
- [, J Brackets denote data known to the staff and considered proprietary to i
i the applicant and specified in reference 6 to this report.
l I
l 2-6 power shapes to determine the axial location at which the decrease in the departure from nucleate boiling ratio (DNBR) due to the superimposed power spike is maximized. These calculations also include the increase in average linear heat generation rate from 6.11 kW/ft to 6.25 kW/ft due to the reduced fuel column height based on the instantaneous densification from the minimum initial density of [ ] theoretical density (TD) to a final density of 0.965 TD.(I) The reactor operating limits, which will be part of the Technical Specifications for Rancho Seco, Unit 1, are based on maximum linear heat generation rate through the reactor power vs axial imbalance correlation.
2.3.2 Fuel Rod Thermal Analysis The applicant uses the B&W computer code, TAFY(II), to calcu-late gap conductance, fuel temperature, and stored energy for the
~
Rancho Seco Unit 1 fuel, which in turn are used in the safety analyses. To demonstrate the applicability of the TAFY code for the evaluation of the Rancho Seco Unit 1 fuel thermal behavior, the applicant compared TAFY predicted fuel temperatures and gap conductance with experimental data.
The staff reviewed the TAFY code and concluded that realistic and/or conservative assumptions have been used for the modeling of the physical phenomena incorporated into the code (thermal expansion, fuel swelling, sorbed gas release, fission gas release),
with two exceptions: (1) partial contact between the clad and fuel and (2) formation of a central void due to fuel restructuring on h
I
~
2-7 the basis of columnar grain growth at a temperature of 2300'F.
Details of the staff's evaluation of the TAFY code and its applica-tion to Rancho Seco Unit I fuel rods are given in the generic evaluation.(7)
Because of the two exceptions noted above, the staff re-quired the applicant to analyse the fuel thennal performance using a 25% reduction in gap conductance and taking nc credit for fuel restructuring. This analysis (6) resulted in a reduction in the peak linear heat rate at which centerline fuel melting would occur from 22.2 kW/ft before densification to 20.4 kW/ft after densification was conservatively taken into account. The reactor protection system prevents fuel centerline melting from occurring for all steady state anticipated transients. This is accomplished by proper setting of the reactor trips as a function of power level and axial power imbalance. These settings will be specified in the Technical Specifications.
2.3.3 Steady State and Loss-of-Flow Transient The effect of fuel densification on DNBR during steady state and design overpower operation was analyzed by both the applicant and the staff.I7) The results show that the steady state minimum DNBR decreases due to an increase in the surface heat flux resulting from fuel densification. To assess the amount of reduction nn DNBR margin, the applicant reanalyzed the steady state operating and design w.
D g
p y
r
.c g
2-8 overpower conditions with an assumed axial power shape that peaked near the core outlet rather than with the symmetrical reference design power shape described in,the FSAR. The outlet shape, though not achievable in operation, produces the largest possible DNBR penalty from fuel densification, because the point of minimum DNBR is shifted toward the top cf the hot fuel rod where the densification induced power spike is the largest. The application of this large power spike at the point of minimum DNBR produces the greatest degradation in DNBR. Using this outlet axial power peak, the applicant computed a 4.18% reduction in DNBR from the 1.39 value reported in the FSAR at overpower without the effects of densification. The applicant his proposed more stringent control rod positions and axial offset limits to compen-sate for the loss in DNBR margin. This is acceptable to the staff.
B&W also reanalyzed the loss of flow transient that would result from a loss of electrical power to the reactor coolant pumps taking into account the effects of fuel densification. The
]
results show that the minimum DNBR during the transient decreased i
due to local flux increases caused by fuel densification. The l
previously calculated minimum DNBR during the transient was 1.45 whereas with the densification the minimum DNBR is calculated to be approximately 1.40.
i i
~
l 1
2-9 The dent fication effects that could aggravate the con-i sequences of the loss-of-flow transient are the increase in the steady state fuel temperature (stored energy), increase in heat flux, and a decrease in gap conductance. The increase in fuel temperature provides more stored heat in the fuel which must be removed during the transient; the higher heat flux provides greater initial enthalpy in the coolant channel. The decrease in gap conductance delays the removal of heat from the fuel resulting in a higher ratio of heat flux to channel flow dering the transient and thus a lower DNBR.
The reduction in DNBR aue to the effects of fuel densification has been evaluated and found acceptable.
2.3.4 Other Transients The following other transients have been reviewed to detennine whether the effects of densification have resulted in significant changes in their consequences:
(1) Control Rod Withdrawal Incident (2) Moderator Dilution Incident (3) Control Rod Drop Incident (4) Startup of an Inactive Reactor Coolant Loop (5) Loss of Electrical Power In the applicant's FSAR these transients were calculated to result in a DNBR in excess of 1.3, or their consequences were shown to 5.e limited to acceptable values by limits to be set forth i
t w
~
2-10 in the Technical Specifications. The staff has reviewed these transients taking into account the effects of fuel densification and agrees with the applicant that they would not result in a reduction of the core thermal margin, i.e., a DNBR less than 1.3.
2.3.5 Sumary The effects of fuel densification on steady state and transient operation have been evaluated by the applicant and reviewed by the staff.
The effect on steady state operation, mostly due to local increases in thennal neutron flux and heat generation, is to require more restrictive limits on control rod positions and offset limits in the Technical Specifications for Rancho Seco Unit 1.
In order to prevent fuel melting, the maximum allowable linear heat generation ratt at 112% overpower has been reduced from 22.2 kW/ft to 20.4 kW/ft.
The staff concluded on the basis of its review that the potential effects of fuel densification on steady state and postulated transient operation have been evaluated in an appropriate manner and are acceptable for the type of fuel fabricated for the first fuel cycle.
2.4 Accident Analyses 2.4.1 General Analyses of the consequences of various postulated accidents were presented in the FSAR for the Rancho Seco Unit 1.
The accidents evaluated were:
2-11 (1) Locked Rotor (2) Loss-of-Coolant (LOCA)
(3) Control Rod Ejection (4) Steam Line Rupture (5) Steam Generator Tube Rupture (6) Fuel Handling (7) Waste Gas Tank Rupture Since fuel densification will affect the consequences of the first four postulated accidents they have been reanalyzed by the appifcant and reevaluated by the staff. Results of the first three accidents (locked rotor, loss-of,oolant, and con-trol rod ejection) are presented in separate parts of this section. The steam generator tube rupture, waste gas tank rupture, fuel handling and steam line rupture accidents are discussed below.
Changes in the fuel pellet geometry can cause the stored energy is the fuel pellet to increase by the mechanisms discussed in Section 2.3 of this report. Potential increases in local power due to the formation of axial gaps are discussed in Section 2.3.1 Both of these effects are accounted for in the evaluation of accidents.
The radiological consequences of accidents were independently calculated by the staff. The results of the staff's calculation of the radiological consequences of accidents were presented in the W
e-,
--4
-g-y
t 2-12 Rancho Seco Unit 1 Safety Evaluation Reportdated June 8,1973.
The radiological consequences would not increase as a result of fuel densification, although the transient performance of the fuel rods can change as a result of fuel densification. It is the latter factor that is discussed in the following sections.
The staff evaluation of the radiological consequences of a waste gas decay tank failure was based on an assumed quantity of gas in the tank which is limited by the Technical Specifications.
For the steam generator tube rupture accident, the assumed quantity of reactor coolant activity is consistent with the Technical Specification limits on maximum permitted reactor coolant system activity. Fuel densification will not affect the conse-quences of these accidents.
The postulated refueling accident assumes the dropping of a.
fuel assembly in the spent fuel pool or transfer canal. The fuel rods are assumed to be at approximately ambient temperature during the postulated accident. Therefore, the direct effects of fuel densification will not a7fect the consequences of this postulated accident. The potential for mechanical failure of a flattened rod might be different from that of a normal rod; however, since the staff evaluation has been based on the conclusion that no clad collapse will occur during the fuel cycle (Section 2.2), this potential change in fuel rod characteristics was not considered.
Furthermore, all of the rods in the dropped assembly are assumed to fail.
2-13 The steam line break accident was analyzed by the appli-cant in the FSAR without the effects of fuel densification.
That analysis showed that the worst consequences from this accident would result at the.nd of life (EOL) of the core.
Since the DNBR margin is higher at the EOL, including the effects of fuel densification, the staff does not expect that the thermal limits will be more severe than those presented in the FSAR.
2.4.2 Locked Rotor Accident The reactor coolant system for Rancho Seco Unit 1 consists of two loops; each return from the steam-generator to the reactor consists of two cold legs, i.e., a total of four reactor coolant pumps are used. Locked rotor accidents are characteristically less severe for 4 pump plants than for 3 or 2 pump plants.
The analysis of the locked rotor accident was originally presented in Section 14 of the FSAR. The transient behavior was analyzed by postulating an instantaneous seizure of one reactor pump rotor. The reactor coolant flow would decrease rapidly and a reactor trip would occur as a result of a high power-to-flow signal. The core flow would reduce to about three fourths its normal full-flow value within two seconds. The temperature of the reactor coolant would increase, causing fluid expansion with a resultant pressure transient which would reach a peak of approxi-mately 40 psi above nominal. The applicant computed a maximum cladding temperature of 1390 F at about 3.5 seconds for this accident.
h
.-y
2-14 The staff perfomed independent calculations for this postulated accident using Oconee 1 parameters. The results of these calculations.57) showed that calculated cladding temperatures varied from 670*F to 1720'F, all acceptably low, even with conser-vative variation of input parameters.
2.4.3 LOCA Analysis The B&W evaluation model described in the AEC Interim Acceptance Criteria and Amendments for Emergency Cure Cooling Systems was used by the applicant to evaluate the loss-of-coolant accident (LOCA) for Rancho Seco Unit 1.
The analysis was performed with the CRAFT code for the blowdown period and the THETA code for the fuel rod heat up. The applicant's LOCA analysis without the assumption of fuel densification is reported in the Rancho Seco Unit 1 FSAR 2
based on the 8.55 ft split break in the cold leg at the pump discharge as the limiting break size and location.(12)
During the blowdown period, the gap conductance, reduced due to fuel densification according to the staff requirements, could cause the core average fuel pellet temperature to increase, but CRAFT calculations show that the average temperature experiences only a very small change. Since in the initial analysis an average core temperature was used that is higher than the average core temper-ature resulting from the decreased gap conductance, the applicant con-1 cludes that the limiting break size and locations do not change due to fuel densification.
i
2-15 The effects of fuel densification on the reficod calcu-lations are small, since the gap conductance is much larger than the film coefficient (cladding surface to coolant) during reflood. The film coefficient is thus limiting with regard to heat transfer and cladding temperature.
The applicant originally performed the LOCA analysis in BAW-1393(0) with an axial power shape that peaks [ ] above the core mid-plane and a corresponding axial peaking factor of F = 1.497 which includes an axial uncertainty factor of 1.024 and a local factor of 1.026 accounting for the effect of the grid structure on the axial peak. This particular flux shape results in the highest linear heat rate and occurs during the control rod maneuvering resulting from the 4-day design basis transient. The design basis transient is defined as a 100%-50%-100% transient, consisting of operation at 100%
power, reduction to 50% power, operation at 50% power for about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and return to 100% power.
THE THETA calculations were performed with the staff require-ments for initial fuel pellet density assumptions. However, instead of imposing a power spike due to a fuel column gap at the peak axial power [ ] above core midplane the applicant used an equivalent radial multiplier over the entire length of the fuel pin which leads to a higher calculated peak cladding temperature f
i
r a
2-16 of approximately 10*F. A hot channel factor of FHC = 1.014 was used in the calculations. The radial peaking factor, F '
R including an uncertainty factor of 1.05, was varied until the calculated maximum cladding temperature approached the 2300*F limit. Using the gap conductance, as calculated with the TAFY code described in Section 2.3.2 of this report, a clad temperature of 2286*F was reached with a maximum allowable linear heat rate of 19.1 kW/ft. The applicant has submitted additional information(13) which further reduces the maximum allowable linear heat rate from 19.1 kW/ft to 18.2 kW/ft due to use of an axial peaking factor of 1.7 versus the 1.497 value previously used. The sensitivity of the core axial power distribution and the moderator temperature coefficient on the analysis of the postulated LOCA is discussed in Sections 2.4.3.1 and 2.4.3.2 of this report.
2.4.3.1 Core Axial Flux Profile The applicant has submitted an analysis in a letter report,03)
TRG-73-47, on Operation Parameters for Rancho Seco Unit 1, dated October 1973, which provides the sensitivity of axial power distribution on ECCS-performance. This analysis assumed that the peak fuel rod power was located at various axial elevations, and determined the maximum linear heat rate (kW/ft) that would result in a' calculated cladding temperature near the 2300*F limit for
-the postulated limiting LOCA. An axial peaking factor of 1.7 was assumed in these calculations. The maximum allowable linear heat f
e v-y
2-17 rate is reduced from 18.2 kW/ft to approximately 15 kW/ft as the location of the peak is moved from the 6 ft elevation tc the 10 ft elevation.
The applicant has proposed to operate the Rancho Seco reactor such that the limiting heat rate as a function of axial elevation as provided in Figure 4-1 of the letter report (13) is not exceeded.
The staff agrees with this approach, and will incorporate the necessary requirements in the Technical Specifications.
2.4.3.2 Moderator Temperature Coefficient The applicant has performed analysesO3) showing the sensitivity of the maximum allowable heat rate during the LOCA to changes in the moderator temperature coefficient. Moderator temperature coefficients in the range from -1.8 X 10~4 to +0.9 X 10-4 Ak/k/*F snow a reduction in the linear heat rate of 0.6 kW/ft; further, in the negative rar ge from 0.0 to -1.8 X 10~4 ak/k/ F the reduction in the 0
linear heat rate is 0.2 kW/ft. The LOCA analysis was performed with a zero coefficient. The applicant has shown (13) that at or below 95% of rated power the Interim Acceptance Criteria will not be exceeded with a positive moderator temperature coefficient of +
0.9 x 10-4 ak/k/ F which bounds the maximum predicted value. The j
Technical Specifications will, therefore, prohibit operation above 95% of full licensed power unless the moderator temoerature coefficient is zero or negative.
2.4.4 Rod Ejection Accident The control rod ejection transient has been reanalyzed (6,14) by the applicant to account for changes in the fuel due to densification.
The significant effects of fuel densification are an increase in the initial maximum fuel temperature and a slioht increase in
~
2-18 4
-average heat flux due to shrinkage of the pellet stack length.
In addition, spikes in the neutron power can occur due to gaps in the fuel. Calculations have verified that no changes in the basic kinetic response of the core occur due to the small changes in fuel geometry and heat transfer characteristics.
The results of the rod ejection accident at beginning of life (BOL) and at EOL without consideration of densification effects have been previously presented in the Rancho Seco Unit 1 FSAR. The staff consultants at Brookhaven National Laboratory (BNL) have performed independent check calculations using appropriate input data and their own computer codes and have confirmed that the results of a rod ejection transient are less severe at EOL than at BOL. Thera-fore, all calculations by the applicant considering densification effects were done for BOL conditions.
For.the full power transient, the control rod reactivity 4
worths available for the assumed ejected rod would be expected to I
decrease because of the more restrictive insertion limits on the control bank. However this was not included in the reevaluation, thereby adding additional conservatism to the calculations. The maximum Technical Specification rod worth of 0.65% a k/k was used for the BOL calculations at full power.
The staff review of the initial fuel temperature for the BOL full power case indicated that a reasonable temperature was used for i
the assumed conditions, consistent with that used in the LOCA i
w r-yr-
-mw-gr---
ewyw--y w
y
-r
- =
4 2-19 analysis. The assumption of no clad creepdown is conservative in that creepdown will reduce the gap which increases gap conductance.
The neutron power spike effect was included in the reanalysis.
The reexamination of the rod ejection transient considering the effects of densification has resulted in a peak pellet average enthalpy of 230 cal /gm which is well below the staff's criterion of 280 cal /gm. The maximum clad temperature reached during the transient is 1580*F. Thirty-three percent of the fuel pins were calculated to be in DNB. The staff review of the rod ejection analysis indicates that reasonably conservative consid-eration has been given to the effects of fuel densification and that the results are acceptable for this accident.
2.5 Summary and Conclusions The effects of fuel densification have been considered in analyses of normal operation, operation during transient conditions, and postulated accident conditions. On the basis of the staff review of the applicant's calculations, and independent calculations performed by the staff and its consultants, the staff concluded that for the period of operation, namely the first fuel cycle:
(1) The effects of densification during steady state and transient operation of the Rancho Seco Unit I reactor will not cause the values on DNBR, and centerline temperatures, to become less conservative than values 1
l
2-20 l
previously established in the FSAR.
(2) The effects of densification were included in the calculation of fuel rod behavior during postulated loss-of-coolant accidents. The LOCA analysis is acceptable and complies with the Interim Acceptance Criteria.
(3) The applicant's omission of the creepdown effect, which tends to increase gap conductance with life time is acceptable.
(4) The Technical Specifications will limit the fuel residence time to 12,000 effective full power hours of power operation to assure no cladding collapse.
(5) The applicant has adopted the staff recommendations for calculating gap conductances and fuel temperatures (Section 2.3.2) as they are used in steady state, transient and accident conditions.
(6) Operating restrictions as necessary to assure compliance with items (1) through (4) above, will be incorporated into the Technical Specifications.
On the basis of the above summary, the staff concludes that the applicant is in compliance with the staff densification reportII) and that Rancho Seco Unit I reactor can be operated at power levels
=-
2-21 up to 100% of rated power with no undue risk to the health and safety of the public.
Operation beyond the first fuel cycle has not been defined at this time. Means that permit safe operation beyond tha first fuel cycle that the staff could find acceptable include: (1)use of a fresh load of the same fuel, (2) continued use of the first cycle fuel with suitable allowance for cladding collapse if pre-dicted, or (3) partial loading with fuel of improved design. Other methods may be proposed, e.g., operation at reduced system pressure.
In sumary, with respect to the matter of fuel densifi-cation after the first fuel cycle, means exist for assuring continued safe operation of the plant.
)
l p.
y e-
D
~
2-22 2.6 References 1.
" Technical Report on Fuel Densification of Light Water Reactor Fuels," Regulatory staff, U.S. Atomic Energy Consnission, November 14, 1972.
2.
" Fuel Densification Report," BAW-10054 Topical Report (Proprietary),
January 1973 (Nonproprietary Information in BAW-10055).
3.
"0conee 1 Fuel Densification Report," BAW-1387 (Proprietary),
January 1973 (Nonproprietary Information in BAW-1388).
4.
" Fuel Densification Report," BAW-10054 - Rev.1 Topical Report (Proprietary), April 1973.
5.
"0conee 1 Fuel Densification Report," BAW-1387 - Rev.1 (Proprietary),
April 1973. (Nonproprietary Information in BAW-1388).
6.
" Rancho Seco Unit No.1 Fuel Densification Report," BAW-1393 (Proprietary), June 1973 (Nonproprietary Information in BAW-1394).
7.
" Technical Report on Densification of Babcock & Wilcox Reactor Fuels,"' by the Regulatory staff, U.S. Atomic Energy Commission, July 6,1973.
8.
Letter from R. C. DeYoung to R. Edwards, Babcock & Wilcox dated April 23,1973, with copy to Sacramento Municipal Utility District.
9.
Peaking Factors in Pressurized Water Reactors with Fuel Den-i sification; BNL Interim Report, December 1972.
i
l i
I
2-23 l
l t
- 11. "TAFY-Fuel Pin Temperature and Gas Pressure Analysis,"
BAW-10044, Topical Report, April 1972.
- 12. "Multinode Analysis of B&W's 2568 MWt Nuclear Plants During a Loss-of-Coolant-Accident, "BAW-10034, Rev.1. May 1972.
- 13. " Operation Parameters for Rancho Seco, Unit 1" TRG-73-47, October 1973.
- 14. " Fuel Densification Report," BAW-10054 Rev. 2 Topical Report (Proprietary), May 1973, (Nonproprietary Information in BAW-10055).
e s+'
b
-***w c
y y-
--.-3---
...v,,,
w
3-1 3.0 LARGE, SMALL AND CORE FLOODING TANK LINE BREAKS - LOCA's The staff's evaluation of the available information in re-gard to the limiting large line, small line, and core flooding tank (CFT) line breaks was presented in the Rancho Seco Unit 1 Safety Evaluation Report dated June 8, 1973. This evaluation is updated in the following sections to include information requested by the staffb)* specific to Rancho Seco Unit 1.
This information does not include the effects of fuel densification. The effects of fuel densification on the LOCA analysis are discussed in Section 2 of this supplement.
3.1 Large Break Analysis The staff reviewed the applicant's reflooding analysis using a new carryover rate fraction ** correlation develooed by B&W during the course of the rulemaking hearing (Docket RM50-1) to account for the entrainment of reflooding water. The previous reflood analysis performed by B&W and described in Section 14 of the FSAR used an entrainment assumption of 20% of the inlet core flow rate. The 20%
entrainment assumption was based on data obtained from the FLECHT***
program. The staff requested a reanalysis of the reflooding transient using the new CRF correlationU)
Because the new carryover rate fraction correlation took many FLECHT experiemntal runs at different conditions into account, the staff viewed it as a better approach in calculating reflooding rates.
- Numbers in () refer to references listed in Section 3.5.
- The carryover rate fraction (CRF) is defined as the total core mass flow rate out of the top of the core divided by the total mass flow into the bottom
.i of the core.
- Full Length Emergency Core Heatup Tests, t
,e
3-2 The staff reviewed the B&W reflood code (REFLOOD) and compared its results with those of the FLOOD l ' code (an ANC/AEC reflood program) for a typical vent-valve plant. Reflooding rates predicted by both computer programs agree to within 1%,
when the REFLOOD code uses the new carryover rate fraction to predict the entr inment. When the old entrainment assumption of 20% was used, the flooding rates calculated by REFLOOD were significantly higher than those predicted by FLOOD 1.
The applicant recalculated the reflooding rates and heat transfer coefficients for the worst case large reactor coolant cold leg break using the new carryover rate fraction correlation. The heat transfer coefficients used in determining the peak clad temperature were determined from the FLELHT correlation presented in WCAP-7665, (2) with the new, lower reflooding rates.
Peak cladding temperatures calculated using the new reflooding rates were higher, and the cladding remained at elevated temperatures for longer time periods.
A worst case large cold leg break has been analyzed by B&W using the carryover rate fraction entrainment correlation for Rancho Seco at 102% of the rated power level (2772 MWt). These results, presented in Appendix 14A of the FSAR indicated a
3-3 maximum hot spot cladding temperature of 2299'F occurred at 38.5 seconds. Both the maximum clad temperature and the per-centage metal-water reaction calculated using the carryover rate fraction correlation had increased somewhat but were within the limits set forth by the Interim Policy Statement.
The use of the new carryover rate fraction correlation pro-vided a more conservative method of predicting reflood water entrainment than the 20% entrainment assumption since the use of this correlation resulted in lower reflooding rates, higher peak cladding temperatures and greater metal water reactions. The staff concluded that, based on the present experimental data, the use of this more conservative approach was warranted. The staff further concluded that the ECCS performance analysis using this more conservative approach meets the acceptance criteria, as described in the Commission's Interim Policy Statement.
3.2 Small Break Analysis The generic small break analysis for a B&W 2568 MWt plant is presented in BAW 10052(3)
Since the Rancho Seco Unit 1 plant differs only in that it has a higher power level (2772 MWt) the model developed in BAW 10052 is applicable. The worst case small reactor coolant j
2 pipe break (0.04 ft suction break) was evaluated for the Rancho l
Seco Unit 1 plant using the CRAFT computer program. Analysis for this break is presented in Appendix 14A of the FSAR. The higher p%
3-4 power level of Rancho Seco Unit 1 produces enough two phase mixture to keep the core covered. At approximately 2100 seconds, the leak-rate matches the injection rate and the residual ifquid level starts to increase since the injection rate is about constant and the leak rate is decreasing. Temperature calculations were not performed since BAW-10052 shows no cladding temperature increases would be experienced if the core remains covered during the transient.
All conditions of the Interim Acceptance Criteria were met; the peak clad temperature did not increase during the transient 2
0 and was well below 2300 F, there was no metal-water reaction, the core geometry was still coolable and long term cooling was established. The staff evaluation of this updated analysis confirms that the emergency core cooling system will provide adequate protection for small breaks in the reactor coolant system.
3.3 Core Flooding Tank (CFT) Line Break Analysis In response to a staff request (I} to analyze the CFT line break for the Rancho Seco Unit 1 plant, the applicant has submitted an analysis for a 2772 MWt vent valve plant in Amendment 23 to the FSAR. The' analysis used the model reported in B&W Topical I4)
Report BAW-10064
~
3-5 The applicaat's calculations indicate that the ccre was covered by a two phase mixture during the entire transient.
Since the core was covered with a two-phase mixture, pool film boiling provided sufficient cooling and the maximum cladding temperature was calculated to occur at the start of the transient and corresponded to the hot spot steady state value. This analysis showed the cladding temperature resulted in no metal water reaction, core geometry remained unchanged and would not be detrimental to core cooling and decay heat was removed for an extended period of time. The staff evaluation of these analyses further confirms our prior conclusions that the Rancho Seco emergency core cooling system will perform acceptably if needed.
3.4 Cnnelusions Based on the evaluation of the applicant's reanalyses described above, and of the modifications to these analyses that have been made to include the effects of fuel densificaticn (as reported in Section 2 of this report), we conclude that the emergency core cooling system is acceptable and provides adequate protection for any LOCA, and satisfies the Commission's Interim Acceptance Criteria.
W
-4 3-6 3.5 References 1.
Letter from A. Schwencer to E. X. Davis, Sacramento Municipal Utility District dated April 30, 1973.
2.
"PWR FLECHT Final Report," WCAP-7665, dated April 1971.
3.
"Multinode Analysis of Small Breaks for B&W's 2568 MWt Nuclear Plants, "BAW-10052, September 1972.
4.
"Multinode Analysis of Core Flooding Line Break for B&W 2568 MWt Internal Vent Valve Plants," BAW-10064, April 1973.
t 6
4-1 4.0 OTHER MATTERS 4.1 Diesel Generator Schematic Review During the confirmatory review of the diesel generator control and alarm drawings, we noted that a manual pushbutton was provided on the diesel generator local control panel for emergency engine shutdown purposes. We also noted that this manual pushbutton'would override automatic emergency start of the engine. We required that the design be modified so that the presence of a safety feature actuation signal would remove this capability for manually shutting the engine down from the local control panel. The applicant agreed to make this modification in the design and has submitted the appropriate drawing showing an acceptable implementation for this modification.
4.2 Emergency Feedwater System The staff identified failures of single active components coincident with certain postulated emergency feedwater line ruptures that could cause loss of this vital system. This matter has been discussed with the applicant who has agreed to make the necessary modifications to correct the situation.
The staff will review the details of these modifications for acceptability prior to issuance of a license.
=
5-1 5.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The ACRS has issued a report on its review of the Rancho Seco Unit 1 operating license application. A copy of this report is attached as Appendix B.
The staff has considered the comments and reconnendations contained in that report. The steps which the staff has taken or will take relative to these comments and recommendations are described in the following paragraphs.
5.1 Operation at 2772 MWt The Committee recommended that three conditions be satisfied before this plant be allowed to operate at full power.
The operation of Unit 1 of the Oconee Nuclear Station a.
should be reviewed and found satisfactory by the kegulatory staff.
b.
Following an appropriate period of operation at power levels up to 2568 MWt, the operating experience of Rancho Seco Unit 1 should be reviewed by the Regulatory staff and the ACRS.
Prior to the review in b above, the Regulatory staff c.
should p' rform and report on an independent confirmation e
of the applicant's linear heat generation rates, operating limits and ECCS efficacy.
The staff expects to complete its review of the Oconee 1 power operation and the Rancho Seco Unit 1 operation at 1568 MWt before Rancho Seco Unit 1 is ready for operation at full licensed power and will discuss our evaluation of the operating oerformance of both facilities with the ACRS.
4'
5-2 With respect to c above, the staff is developing techniques for analyzing ECCS performance in anticipation of promulgation of new ECCS criteria. The next step after development of these techniques is to apply them to different classes of plants (i.e., classes arranged by vendor and subclasses within a given vendor's product lines) which would provide independent confirmation of analyses, in a generic sense.
That is, where a series of plants are alike, we would study and confirm only one. At present, the staff has not selected the base-case plants that it intends to use for these independent confirmation studies.
Even if Rancho Seco is not selected for this purpose, a plant sufficiently similar will be analyzed so that the intent of the ACRS recommendation will be fulfilled.
5.2 Fuel Loading Procedures Detailed fuel loading procedures will be developed by the applicant, which provide for obtaining a permanent pictorial record of the installed location of every fuel assembly. These procedures and pictorial record will be reviewed by the staff.
Fuel loading restrictions will be incorporated into the Technical Specifications to assure that each preclassified fuel assembly will be located in a specific predetermined core region.
5.3 Tritium Management The Comittee believes that the effects of gradual buildup of tritium in liquids within the plant should be carefully evaluated.
Based on expected leakage and make-up water, the applicant is expected to be able to keep the plant tritium at reasonable and acceptable levels. To assure that this is done, the staff will i
l 5-3 monitor the semiannual report required by the Technical Specifications for tritium buildup in plant liquids.
If the buildup rate exceeds expectations, the staff will take appropriate action to require the plant tritium inventory to be reduced.
5.4 Positive Moderator Temperature Coefficient The Committee recomended that the matter of limitations applicable to operation with a positive moderator temperature coefficient will be restricted in the Technical Specifications to values less than (more negative) those values employed in the safety evaluation accident analysis. Further, operation above 95% power will be prohibited unless the moderator temperature coefficient in zero or negative. Section 2.4.3.2 of this report provides additional information on this matter.
5.5 Pump Overspeed The staff is investigating on a generic basis the probability and the consequences of rupture of a reactor coolant pipe which in certain locations could result in reactor coolant pump overspeed.
If this study indicates that additional protective measures are warranted to prevent significant pump overspeed or limit the potential consequences to safety related equipment, the staff will require the applicant to provide these protective measures.
5.6 Comon Mode Failure and Anticipated Transients Without Scram The staff has completed its generic review of anticipated transients without scram (I)* (ATWS) and has notified the applicant
- Numbers in ( ) refer to references listed in Section 511 W
e 5-4 of the steps to be taken to minimize the consequences of The applicant has been requested (2) transients without scram.
to submit a detailed report describing the actions that will be taken regarding this matter for the staff's review and evaluation.
5.7 Course-of-Accident Instrumentation The ACRS suggested that the applicant assure itself that instrumentation for determining the course of potentially serious accidents, on a time scale that will permit appropriate emergency action, is provided at the station.
In addition, the applicant was asked to assure that appropriate calibration methods and calculated bases for interpreting instrument responses are available. The applicant stated assurance with regard to these matters in a letter to the staff dated October 24, 1973.
The Regulatory staff, in the operating license review, found Rancho Seco, Unit I acceptable with respect to course-of-accident instrumentation for the following reasons:
a.
Safety related instrumentation in the reactor building is qualified to operate in the post accident environment.
b.
The status of engineered safety features is displayed in the control room so that the operator can verify system operation or take prompt corrective action where necessary.
c.
The calculated responses to major accidents, which will be included in the applicant's accident evaluation program, enable a trained operator to judge the adequacy of system response after an accident.
l
\\
5-5 The Regulatory staff is presently reviewing this matter with the expectation of publishing a Regulatory Guide which would give specific guidelines for instrumentation and emergency procedure preparation.
5.8 Applicant's Management Safety Review Committee The ACRS recommended in its September 11, 1973 letter on Rancho Seco that noncorporate voting members be included in the Rancho Seco Management Safety Review Committee (MSRC). Regulatory Guide 1.33, (Formerly Safety Guide 33), dated November 3,1972, has adopted as acceptable Regulatory practice ANSI Standard N.18.7, which specifies the conditions for which outside consultants as members of a committee such as the MSRC would be needed.
The applicant and the staff have reevaluated the qualifications of the MSRC members and have concluded that additional expertise is required in the area of metallurgy. The applicant has agreed to add a qualified metallurgist to the MSRC as a voting member.
The applicant is also required, by Technical Specifications, to use outside consultants as voting members in all matters for which the owner organization lacks the required specialized training or experience.
We have concluded that the applicant's MSRC when augmented with a qualified metallurgist will satisfy the requirements of Regulatory Guide 1.33 and is, therefore, acceptdale.
M e
--w
~
5-6 5.9 Control of Power Peaking Factors and Linear Heat Rate The Committee recomended that the staff establish suitable criteria for those measures which will be taken to prevent operating under conditions which might result in exceeding acceptable fuel limits established from accident studies and other considerations. The Comittee also recommended that the staff provide suitable bases for evaluating future core loadings (beyond the first fuel cycle).
The applicant will be required to provide alarms and admini-strative procedures acceptable to the staff prior to operation of Rancho Seco, Unit 1 to prevent exceeding acceptable fuel limits.
In addition, power distribution maps will be required periodically during steady state and following transient operation in order to verify predicted power distributions. The applicant will submit a report on the operating performance of the first fuel cycle.
The staff will perform its usual evaluation of this report to establish the bases for future fuel loadings.
5.10 Changes in AEC ECCS Acceptance Criteria In the event of changes in the AEC ECCS Acceptance Criteria, operating limits will be reevaluated and any required changes will be incorporated into the Technical Specifications.
l
g 5-7 5.11.
References 1.
" Technical Report on Anticipated Transients Without Scram for Water Cooled Reactors", Regulatory Staff, U. S. Atomic Energy Commission, September, 1973.
2.
Letter from A. Giambusso to E. K. Davis, Sacramento Municipal Utility District, dated October 9, 1973.
... ~...
m
-%y e
f4 re--
-gr
--w.
p w
r-,r-g v-=-m--
e y
T-9 9
N-
6-1
6.0 CONCLUSION
S The staff's conclusions as stated in the SER remain unchanged. We will, however, restrict initial operation to power levels up to 2568 MWt. We will approve operation at 2772 MWt only after, (1) the Oconee, Unit 1, prototype plant operating perfomance has been found to be satisfactory by the Regulatory staff, and (2) our review of the Rancho Seco, Unit 1, startup test reports and initial operating perfomance at 2568 MWt has been completed and has been discussed with the ACRS.
4 4
[
m-
N A-1 SUPPLEMENT TO THE CHRONOLOGY OF THE SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION UNIT 1 1.
June 4, 1973 Submittal of Amendment No. 22, containing revisions and additional pages to the FSAR.
2.
June 5,1973 Letter from applicant regarding meteorological data collection system.
3.
June 8, 1973 Letter to JCAE transmitting Safety Evaluation dated June 8, 1973.
4.
June 8,1973 Letter to applicant transmitting Safety Evaluation dated June 8, 1973.
5.
July 9,1973 Letter from applicant transmitting report entitled
" Rancho Seco Nuclear Service Spray Ponds Performance Evaluation".
6.
July 25,1973 Letter from applicant requesting withholding from public disclosure and transmitting BAW-1393, " Rancho Seco, Unit 1 Fuel Densification".
7.
August 1, 1973 Letter from applicant transmitting R0 Bulletin 73-2.
8.
August 6, 1973 Letter from applicant regarding July 31, 1973 meeting submitting changes to FSAR and the control scheme for emergency diesel engines.
9.
August 14, 1973 Letter to applicant regarding their August 3,1973 letter with attached report pursuant to 10 CFR 50.55(e).
- 10. August 28, 1973 Letter from applicant transmitting the proposed inspection program-main steam and main feedwater system report.
- 11. August 23, 1973 Letter to applicant transmitting amendment to 10 CFR Part 50 and 55.
l i
.- 12. September 12, 1973 Letter from ACRS to Chairman Ray regarding Report on Rancho Seco Nuclear Generating Station.
- 13. September 13,.1973 Letter from applicant transmittinq copies of ACRS letter dated September 9, 1973.
- 14. September 13, 1973 Letter to JCAE transmitting ACRS letter dated September 11, 1973.
18 September 13, 1973 Letter to appifcant requesting additional information.
- 16. September 18, 1973 Letter to Board Member transmitting report dated September 11, 1973 by ACRS on Rancho Seco.
- 17. September 20, 1973 Letter from applicant furnishing information regarding design review of the Borated Water Storage Tanks at Rancho Seco.
- 18. October 3, 1973 Letter from applicant transmitting Babcock & Wilcox Proprietary Report TRG-73-47 entitled, " Operational Parameters for Rancho Seco, Unit 1".
J e
s
- i e
\\
B-1 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIO ENERGY COMMISSION WASH NGToN. D.C. 20545 l
1 September 11, 1973 i
l 1
i Honorable Dixy Lee Ray Chairman U. S. Atomic Energy Coc: mission Washington, D. C.
20545
Subject:
REPORT ON RANCHO SECO NUCLEAR CENERATIEG STATION, UNIT 1
Dear Dr. Ray:
During its 161st meeting, September 6-8, 1973, the Advisory Committee on Reactor Safeguards reviewed the application of the Sacramento Municipal Utility District for a license to operate the Rancho Seco Nuclear Gener-ating Station, Unit 1, at power levels up to 2772 W(t). This project had been considered previously during the 159th meeting of the ACRS, July 12-14, 1973, by Subcomittee meetings in Sacramento, California, on June 13 and 14, 1973, subsequent to a tour of the site, and in Washington, D.
C., on Auguet 22, 1973. In the course of its review, the Committee had the benetit of discussions with representatives and -
consultants of the Sacramento Municipal Utility District, the Babcock and Wilcox Company, the Bechtel Corporation, and the AEC Regulatory Staff, and of the documenta listed. The Comittee last reported to the commission on the construction of this plant in its letter of July,19, 1968.
The Rancho Seco Nuclear Generating Station is located about 25 miles southea'at of Sacramento, California. Uater for this plant will be supplied from the Folsom South Canal. An on-site reservoir will have a capacity of 2500 acre-feet, and two spray ponds can provide cooling water for decay heat removal for about 30 days.
The Rancho Seco nucisar steam supply system employs a Babcock and Wilcox two-loop, pressurized water reactor essentially identical in design to ths Oconee Nuclear Station, Unit No.1, previously reported on by the Co=iittee. However, Rancho Seco will operate at approxinately 87. higher power level and will use control of bcron concentracion in the core cool-ing water to aid in reactivity control dating pouar maneuvering.
%=
Gi e'
c!
S m:
E N
1 5E
=r
_. ~
1 Honorable Dixy Lee Ray S ept ember 11, 1973 The application for a construction permit proposed initial operation at power levels up to 2452 FM(t), the same as the construction permit power level of the Oconee Nuclear Station, Unit 1 which employs a similar The saf ety analyses have been completed assuming a power of reactor.
2568 FM(t). The application for an operating license proposed power levels up to 2772 FM(t) and safety studies have been made at this power.
This increase in power is accomplished by utilizing larger primary cool-anc pumps and by increasing the average coolant temperature rise in the 3
The Camnitten believes that review of the operation of Oconee core.
Nuclear Station, Unit 1 by the Regulatory Staff should be completed and satisfactory performance of Oconee Nuclear Station, Unit I should beIn demonstrated before Rancho Seco Unit 1 is operated at full power.
addition, the Committee agrees with the Regulatory Staff that it would
~
be prudent for Rancho Seco Unit 1 to operate at power levels up to 2568 M4(t) for an appropriata time period and for the Staff and the ACRS to review this experience prior to allowing operation at full poder of 2772 MW(t). Independent confirmation by the Regulatory Staf f of the applicant's analyses of linear heat generation rates, operating limits, and ECCS ef ficacy, and submittal of a supplemental Staff Safety Evalua-tion Report should precede this review for operation at full power.
Fust for the reactor has been thermally resintered with the purpose of reducing fuel densification under irradiation; furthermore, the fuel 3
assemblies are being classified according to their maximum allowable linear heat rate and are to be loaded into the reactor according to this classification. This matter should be resolved in a manner satisfactory to the Regulatory Staf f.
The Connittee wishes to be kept informed.
The applicant has stated that, under normal conditions, reactor produced Thta radioactive liquid wastes will not be released to the environment.
stil be accomplished primarily through processing and reuse of liquiis The Committee believes that the renoved from various reactor systems.
effects of gradual buildup of tritium in' liquids within the plant should ba carefully evaluated. Factors to be assessed include potential in-creases in radiation exposures of operating personnel, possible diffi-culties in proper plant maintenance, and the possible influence of increased tritium concentrations on the consequences of unanticipated releases.
During the hot functional testing of Oconee Nuclear Station, Unit I which including re-was conducted in 1972, damage occurred to some components, actor vessel internals. The design improvements made to Oconee Nuclear The Co=mittee Station, Unit I have been made siso to Rancho Seco Unit 1.
balieva1 th2t these changes are acceptable.
4 g
Honorable Dixy Lee Ray
-3 September 11, 1973 The applicant has been responsive to the Committee's recommendation that suitable instrumentation be sought to monitor for loose parts and for vibration; auch instrumentation has been designed and will be util-ized.
The applicant has proposed appropriate operating limitations to be applied if, at any time during operation, the moderator temperature coefficient of reactivity is positive. This matter should be resolved in a manner satisfactory to the Regulatory Staff.
The Regulatory Staff has been investigating on a generic basis the problems associated with a potential reactor coolant pump overspeed in the unlikely event of a particular type of rupture at certain locations in a main coolant pipe. Some additional protectiva measures may be warranted, and this matter should be resolved to the satisfaction of the Regulatory Staff. The Committee wishes to be kept informed.
The Committee reiterates its previous comments on the need for further study of means for preventing common mode failures from negating reactor scram action, and of design features to make tolerable the consequences of failure to scram during anticipated transients. The Committee believes it desirable to expedite these studies and to Laplanent in timely fashion such design modifications as are found to improve significantly the safety of the plant in this regard. The Committee wishes to be kept informed of the resolution of this matter.
The applicant should assure himself that instrumentation for determining the course of potentially serious accidents, on a tLae scale that will permit appropriate emergency action, is provided at the station and that appropriate calibration methods and calculated bases for interpreting instrument responses are available.
In view of the important role of the applicant's Management Saf ety Review Ccanittee in providing continuing reviews, and in updating and implementing safety measures, the ACRS recommends that the thnagement Safety Review Com-mittee include additional experienced personnel from outside the corporate structure as voting members.
The applicant has proposed measures, including alarms and administrative procedures, to prevent operating under conditions which ni ht result in 3
exceeding acceptable fuel limits established frem accident studies and other consideracians. The current reviev has been confin2d to the first fuel cycle, and the analyaes have been based on the as-built fuel.
The ACRS re:ommends that the Raculatory Staff establish suitable criteria for these measures and provida suitabla bases for evaluating fucura loadings.
The Cae:mittee wishes to be kept infm-ed.
M v
/
Honorable Dixy Les Ray September 11, 1913 The Cocnittee recognizes that re-evaluation of operating limits raay be necansary as a result of possible changes in the acceptance criteria for emergency core cooling' systems. The Committee vishes to be kept informed.
Other problems relating to large weter reactors which have been identified by the Regulatory Staff and the ACRS and cited in previous reports should be dealt with appropriately by the Regulatory Staff and the applicant as suitable approaches are developed.
The Advisory Consoittee on Reactor Safeguards believes that, if due regard is given to the items mentioned above, and subject to satisfactory comple-tion of construction and preoperational testing, there is reasonable assurance that Rancho Seco Nuclear Generating Station, Unit I can be operated at power levels up to 2772 W(t) without undue risk to the health and safety of the public.
Sincerely yours, w
H. G. Mangelsdorf Chairman
Attachment:
List of References
~
g Honorable Dixy Lee Ray Septemb r 11, 1973 References 1.
Sacramento Municipal Utility District (ShMD) Safety Analysis Report for Rancho Seco Nuclear Generating Station, Unit 1, Vols. I-V, May, 1971 and Vol. VI, June, 1972 2.
Amendments 6 through 23 to SMUD License Application for Rancho Seco 3.
Letter from E. K. Davis, SMUD, to A. Giambusso, L, dated March 23, 1973, " Final Report on Minor Imperfections Found in Pipe Welds at the Rancho Seco Nuclear Generating Station" 4
Letter from E. K. Davis, SMUD to A. Giambusso, L, dated April 3,1973,
" Interim Report on Fuel Densification" 5.
Letter from E. K. Davis, SMUD, to A. Giambusso, L, dated May 1,1973,
" Interim Report on Effects of Piping Break Outside Containment" 6.
Letter from E. K. Davis, SMUD, to A. Schwencer, L, dated May 3,1973,
" Review of Control Circuits" 7.
Directorate of Licensing Safety Evaluation, June 8,1973 8.
Letter fran H. W. Ibser, Professor of Physics, California State University to M.Libarkin, ACRS, dated June 18, 1973, concerning temperature inversions at Rancho Seco 9.
Babcock and Wilcox Proprietary Report, BMW-1393, " Rancho Seco Unit I l
Fuel Densification Report," June,1973 with supplemental information containini as-built data forwarded by letter from E. K. Davis, SMUD, to A. Giarbusso, L, dated July,23, 1973 l
- 10. Report, " Rancho Seco Nuclear Service Spray Fonds Performance Evaluation," dated June 29, 1973 by the Waste Heat Management Research Project, University of California, Berkeley
- 11. Directorate of Licansing Technical Report on Densification of B&W Reactor Fuel, dated July 6, 1973 12.
Letter from E. K. Davis, SMUD, to A. Giambusso, L, dated August 2,1973, submitting changes to the FSAR, and the control scheme for emergency diesel engines.
e w. 4.
m w