ML20037B181
| ML20037B181 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/20/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 8009080558 | |
| Download: ML20037B181 (17) | |
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c UNITED STATES 1'
NUCLEAR REGULATORY COMMISsl N 3 g W ASHINGTON, D. C. 20555 o
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDfENT NO. 21 TO FACILITY OPERATING LICENSE NO. DPR-2 CONONWEALTH EDISON COMPANY DRESDEN UNIT NO. 1 DOCKET NO. 50-10
1.0 INTRODUCTION
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By letter dated April 11,1977 Commonwealth Edison Company (CECO) proposed to reload Dresden Unit No. I reactor with replacement 6 x 6 fuel assemblf.os of a different manufacture, but similar characteristics to presently used fuel.
Documentation submitted in support of the proposed rel,oad consisted of a report entitled "Dresden Unit No.1 Cycle 11 Reload Licensing Report".())
In response to NRC que ons, further infomati and September 8,1977.gwas provided by letters dated August 3,1977, During our review, we detemined that certain technical specification changes were necessary in order to assure conservative margins of safety in plant-operation.
These changes have been accepted by CECO.
This amendment would modify the license and technical specifications to allow operation of the facility with:
(1) Up to 66 reload assemblies by Exxon Nuclear Company 6 x 6 fuel assemblies ds replacement for existing fuel assemblies fabricated by different manufacturers.
(2) Revised MCHFR limits to assure conservative operation of Cycle 11 with respect to themal hydraulic analyses.
(3)
Incorporation of new MAPLHGR limits to assure continued ' reactor operation that is acceptable with respect to ECCS perfomance criteria.
2.0 EVALUATION 2.1 Nuclear Characteristics Up to 66 6 x 6 reload fuel bundles with an average enrichment of 2.24 weight
% will be loaded throughout the core.
The core contains a total of 464 fuel bundles, so that about 14 percent of the fuel bundles are being replaced for the reload.
The loading pattern consists of the 6 x 6 reload bundles surrounding the central core region by approximately three rings of fresh fuel in a one-in-four tattern.
The data in Reference 1 indicate that the nuclear characteristics of the Cycle 11 core are similar to the previous core.
Thus, tne total control system worth, temperature, and voic ependent 80090-80 f 8 [
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_2 behavior of the reconstituted (core will not differ significantly from those values previously reported for-thet Dresden Unit No. l reactor. The shutdown margin of the reconstituted core will meet the license requirement that the ccre be at least 0.0025 ak subcritical 'in the most reactive operating state with the most reactive rod fully withdrawn. For the Cycle 11 core the calculated minimum shutdown margin for the beginning-of-cycle (BOC) is such s
tnat the final core loading will meet this requirement. The calculational technique used was the same as that used in Cycle 10, which has been verified with experimentally derived shutdown margins for Cycle 6 and Cycle 7.
J" The information presented in Reference 2a indicates that a boron concen-tration of 525 ppm in the moderator will bring the reactor subcritical. by
<C.01ak at 20*C, xenon free. Therefore, the current Technical Specification requiring a minimum of 400 los' of boron in the Standby Liquid Control System (which can produce a 690 ppm concentration in the core) insures that the alternate shutdown requirement of the General Design Criteria is met by the Standby Liquid Co_ntrol System.
Analyses of the reactivity resulting from temporary storage of fresh fuel in the spent fuel storage rack show that the effective multiplication factor, k of the FN-1 fuel as stored in the fuel storage rack is 190 for $bhm,al conditions. Therefore, the XH-1 fuel meets the fuel storage requirement for:Dresden Unit No.1. '
The void, moderator temperature, and Doppler reactivity defects for the XN-1 fuel'and for a typical p'revious. fuel type are given in Table 5.1.6-1 of Reference 1.
The Doppler.defccts for the two fuel types were found to
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The other two defects were be differert by only a;small percentage..
somewhat different; howeve'r, we agree with 'the licensee's conclusion (response to, Question 3,'-Reference 2a) that the small relative number of XN-1 fuel assemblies being introduced into the core for Cycle 11 (14%)
will not significantly alter the overall core-wide reactivity defects.
Based on the findings stated above, we conclude that the nuclear characteristics and performance of the reconstituted Cycle ll core are acceptable, ar.d.that_ they will not differ significantly from that of the previous core cycle, which-wcs acceptable.
2.2 t'echanical Design The XN-1 fuel type to be loaded for Cycle 11 is fabricated by the Exxon Nuclear Company.
It is similar to that of the previously used Type VI-I assemolies, exceot that cladding tnickness has been increased in some of the rods and the spacer design is changed.
Increasing the cladding thickness is expected to increase design conservatism.
Based on V
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3-similarity to the previous fuel, which was accepted and which has demonstrated acceptable performance for several core cycles, and based on acceptable performance of Exxon Nuclear Company fuel of similar design in other reactors, we find the mechanical design of Type XN-1 fuel acceptable.
2.3 Thermal-Hydraul_ic Desian The licensee's application for reload 5I) stated that "no new (transient) or acc,ident) analyses are indicated," based on similarity of the Cycle 11 core to previous cores that have been used successfully in Dresden. Unit No. 1.
However, it was the NRC staff judgment that the licensee should present new transient and accident calculations applicable to-Cycle 11 and based on an acceptable criticai-heat-flux (CHF) correlation, such as XN-2.
This staff judgment was based on the fact that the existing analyses were performed assuming core characteristics of a previous cycle, using analysis techniques different from those currently employed by most other licensees (for example, equating a steady-state overpower reactor state to a transient reactor state) and using the Janssen-Levy critical heat flux correlation which has been shown to be non-conservative for non-uniform axial power shapes..The licensee responded with a commitment;to provide such analyses by November 1,1977.
The new transient and accident analyses will be performed and submitted, and operating restrictions that result from the more conservative approach described below can be relieved following review and approval of the new analyses. The analyses will also be used as a conservative " reference cycle" analysis which can be utilized by the licensee for future reloads.
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After several discussions between the licensee and the NRC staff, an l'
alternate method was proposed to conservatively bound thermal hydraulic performance of the Cycle 11 core during postulated transients and accidents in the interim period between Cycle 11 startup and November 1, 1977, when the.new analyses will be presented.
That alternate method and the bases for i'ts acceptability are presented in the following sections.
2.3.1 Interim Safety Limit FCHFR The licensee has performed a number of sensitivity studies comparing Minitaum Critical Heat Flux Ratio (MCHFR) values determined by the
Janssen-Levy, correlation to Minimum Critical Power Ratio (MCPR) values determined by the XN-2 correlation. These studies were mace by com-paring steady'-state reactor conditions with dif ferent flows, power levels, axial poser shapes, and different inlet enthalpies (i.e.,
. temperature and void content). For each such reactor condition, both the MCHFR and MCPR values were calculated.
Using the comparative J
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results, tre licensee has determined the most conservative credible
" conversion factor" and has used that " conversion factor" to show that if the MCHFR for Dresden Unit No. I as. determined by the Janssen-Levy enrrelation does not go below a value of 1.9, then it conservatively assures that the MCFR for Dresden Unit No. I as determined by the XN-2 correlation does not go below a value of 1.32.
The value of 1.32 (determined by XN-2) has been found by the staff to be an acceptable safety limit MCPR for other BWR Technical Specification safety limits.
Therefore, a Janssen-Levy MCHFR value of 1.9 is an acceptable limit for Dresden Unit No.1.
2.3.3 Interim Operating Limit MCHFR To insure operation of the plant above the safety limit MCHFR, it is necessary to determine the maximum change in MCHFR (AMCHFR) that could occur during an anticipated transient. Adding that AMCHFR to the safety limit MCHFR and recuiring the plant to operate above the resulting sum go below the safety limit MCHFR during a transient.(}hg fuel will not (called the operating limit MCHFR) will assure that In order to determine the maximum credible MCHFR for Dresden Unit No.1, the licensee and the staff reviewed transient analyses of several other BWR's that had used the Janssen-Levy correlation to determine the worst case MCHFR.
In the Turbine Trip Without Bypass transient, which is believed to be the limiting transient for Dresden Unit No.1, the largest AMCHFR in the analyses of other plants was about 0.50.
However, to assure conservatism, the licensee and the staff established.the largest AMCHFR for all transients in other BWR's, not only the one believed to i
be the limiting transient for Dresden Unit No.1.
It was found that this largest value of AMCHFR is 0.59 for the overpower trip. This 2tiCHFR was then adced to the safety limit MCHFR for Dresden Unit No.1.
To account for any plant differences between the Dresden Unit and the other BWR's surveyed, or for inaccuracies in the analyses used to cal-culate the safety limit MCHFR or the 0.59 AMCHFR, a margin of 0.31 was added to the AMCHFR. The total MCHFR operating limit is therefore 2.80, which the NRC staff agrees is an acceptable, conservative-operating limit MCHFR for Dresden Unit No.1.
It is anticipated that the new analyses to be submitted by November 1,1977, will reconfirm the conservatism in this approach and provide a technical basis for establishing a less conservative operating limit.
%.4 Accident Analysis 2.4.1 ECCS Aooendix K Analysis On December 27, 1974, the Atomic Energy Commission issued ?n Order for Modification of License implementing the requirements of 10 CF9 50.46
" Acceptance Criteria and Emergency Core Cooling Systems for Light l
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Water Nuclear Power Reactors." One of the requirements of the Order was that, prior to any license amendment authorizing any core reloading,
...the licensee shall submit a re-evaluation of ECCS cooling performance calculated in accordance with an acceptable evaluation model which confoms to the provisions of 10 CFR Part 50.46."
The Order also required that the evaluation should be accompanied by such proposed changes in Technical Specifications or. license amendments as may be necessary to implement the -
evaluation results.
By a Memorandum and Order issued by the NRC on August 21,1975,(4)
Commonwealth Edison Company was granted an exemption, in force until December 31,1977, "from the requirements of and underlying 10 CFR 50.46 with respect to the design and diversity of emergency systems or the diversity of emergency power sources (for Dresden Unit No.1) but not from the specific perfomance requirements of the FAC. Credit was given f r) 94 In the Order,t certain equipment and for offsite power availability.
the Commission noted that " Commonwealth Edison submitted on November 1, 1974, a preliminary evaluation of the reactor's ability to comply with the FAC, not necessarily including all detail and documentation called for by Appendix K, but nevertheless based on conservative assumptions and providing a conservative assessment of ECCS perfomance."
Recent staff concerns regarding steam effects on core spray distribution (5)(6) were examined to assure that operation during Cycle 11 would be within tg)
ECCS performance requirements of 10 CFR Part 50.46.- The staff requested that Commonwealth Edison define a " spray distribution that conservatively accounts for the steam atmosphere that would exist in the Dresden reactor vessel following a postulated LOCA" ano further, " Based on this spray distribution and conservatively determined heat transfer coefficients, calculate the maximum allowable bundle power and linear heat generation rate that will assure that the performance requirements of Paragraph 50.46 of 10 CFR are not exceeded."
In response, CECO referenced spray flow distribution calculations made previously that were based on single 1;ozzle tests in air. The air tests were used to determine an individual nozzle's spray' distribution. The combined effect (flow) from all nozzles was then detemined for each bundle, based on geometric calculations.
For each of three concentric core regions, CECO selected the minimum flow predicted for any single bundle and divided that minimum flow by a factor of two to account for effects of a steam environment that were not determined in the air tests.
This procecure was utilized to produce a conservative prediction of the l
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minimum spray flow available to any bundle in each of the three core regions, considering effects of the steam environment. The NRC staff agrees with CECO that a factor of two reduction to account for steam effects is conservative for this purpose.
Dresden IJni' No. 1 utilizes 1/2H30 nozzles which produce a large droplet size O! !.arge droplets have been shown gbe affected less by a steam environment than smaller drop-let sprays.
Based on
'or similar type nozzles,ggle nozzle test results in air and steam we believe the factor of two reduction conservatively accounts for steam effects on core spray distribution for the Dresden Unit.
CECO then detennined an assembly maximum power for each of the three regions such that the minimum spray flow predicted as above for that region (considering steam effects) would provide adequate cooling (i.e.,
so that the spray cooling coefficients already assumed in the ECCS-LOCA analyses could be conservatively justified). This was done by examining numerous reports concerning Full Length Emergency Cooling Heat Transfer (FLECHT) experiments, and examining minimum spray flows conservatively predicted to be present in other BWR's of various types with corre-sponding spray cooling coefficients assumed for those reactors.
It was noted that a certain " vaporization flow" can be defined for each fuel assembly in the reactor or in the FLECHT tests, such that vapor-ization of that amount of water will exactly remove the total amount of heat being produced in the bundle at the earliest calculated time of spray initiation following a postulated LOCA.
It was further noted from the FLECHT tests and from conditions present in other operating BWR's that if the minimum spray flow available to a fuel bundle (considering steam effects) is a factor somewhere between 1.3 and 2.0 above the above defined " vaporization flow," then the spray cooling coefficients assumed in the ECCS-LOCA calculations are con-servatively justified. CECO detennined a maximum bundle power for each region such that the minimum predicted spray for any bundle in the region (considering steam effects) is a factor of two above the " vaporization fl ow. " The net effect is an approximate 307, derate in the maximum power allowable from the highest power bundle in the most affected region. We note that the net effect of the procedure described in this paragraph and in the preceding paragraph results in a factor of four between the minimum flow to any bundle (predicted without consideration of steam effects) and the " vaporization flow" of the highest power bundle.
This conservatively accounts for steam effects and is acceptable.
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~he maximum allowable power from any bundle in each of the three core regions (detemined as described above) will be incorporated into the Technical Specifications, and periodic surveillance will be required to insure' that these regional bundle power limits are not exceeded.
On the above basis, we conclude that when Dresden Unit No. I is operated in accordance with the above limits, it will meet perfomance requirements of the F AC based on the analyses perfomec. using the' ECCS models currently employeo by CECO for Dresdgunit No.1, and therefore the original basis for granting the exemption is still valid.
Dresden Unit No. I therefore meets all requirements of 10 CFR 50.46 (except for certain of those require-ments from which they are specifically exempted until December 31, 1977) and operation of Cycle 11 is therefore acceptable from loss-of-coolant accident considerations within the limitations imposed by that exemption.
2.4.2 Steamline Break Accident Steamline break accidents which are postulated to occur inside containment are considered in the ECCS analysis discussed in Section 2.4.1.
The analysis of steamline bregccidents occurring outside containment were presented by the licensee to support operation of previous cycles.
Since results of the steamline-break-outside-of-containment analyses would not be citanged by the reloaded core, the previous analyses are acceptable.
2.4.3 Fuel Loading Error Fuel loading errors are discussed in Reference 2a for a fresh XN-1 6 x 6 fuel bundle placed in an improper location or rotated 90 or 180 degrees in a location near the center of the core. The infomation pre-sented indicates that the worst fuel loading error results in a minimum critical power ratio (MCPR, based on the acceptable XN-2 correlation) of greater than 1.32, which is an acceptable value. The licensee calculated effects for several other misloadings which we believe reasonably bound the worst case misloading error, thereby providing the basis for our conclusion that the misloading error describcd above is the most severe.
2.4.4 Control Rod Drop Accident Analyses of control rod drop accidents were presented by the licenseeU2) to support operation of previous cycles. The relcaced core in this cycle does not significantly change core parameters that would increase the consequences of a control rod drop accident. ile agree that the previous analyses are acceptable.
2.4.5 Fuel Handling Accident Analyses of the fuel handling accident were presented by the licensteO3) to support operation of previous cycles. The results of those analyses
~. would not be significantly altered _by the reloaded core. Therefore, the fuel-handling-accident was not re-reviewed for this reload.
- 2. 5 Overoressure Analysis in Reference 24 the licensee cited Reference 9 wnere the results of an Anticipated Transient Without Scram ( ATWS) analysis is presented.
That ATWS analysis demonstrates that an adequate margin exists below the ASME code allowable essel pressure of 110% of vessel design pressure.
The transient analyzedl9() was loss of condenser vacuum resulting in a-turbine trip without bypass, which is more severe than itSIV closure for Dresden Unit No. I since the MSIV's close much more slowly than the turbine stop valves. This transient was determined to be the most severe overpressure transient, and it _was analyzed without scram. The results indicate that the peak pressure will not exceed the safety limits in ASME Pressure Vessel Code.
Only seven of the existing ten safety valves opened during this analyzed transient; therefore, the transient could be performed with one failed safety valve assumed (as required by the NRC staff) with little or no effect.
Overpressure analyses accepted by the NRC staff on other reload applications have assumed MSIV closure with high neutron flux scram and one failed safety valve.
However, the assumption of no scram in the above ATWS analysis and the margin present (only seven of ten valves opened) represents a conservatism which more than conipensates for any possible slight non-conservatism associated with small differences in nuclear core physics parameters between the previous cycle analyzed and the present Cycle 11 (which core cycles are not significantly different as stated in Section 2.1 above).
For the reasons stated above, the pressure nargin in the above referencec ATWS analysis (presented in Reference 9) is acceptable for Cycle 11.
2.6 Thermal Hydraulic Stability Analysis The thermal hydraulic stability results are presented in reference 2a.
Commonwealth Edison referenced extensive series of tests that were conducted at plant startup in 1S60 and 1961. The tests included stability tests after disturbances caused by rod oscillations, pump trip, and pres-sure and load changes at half power and at full power.
Extensive high void tests were conducted in 1961.
All of the referenced tests (and analyses of those tests) demonstrated the thermal-hydraulic stability In addition.
of the Dresden reactor under a widgvariety of conditions.
the reactor has generated 1.6 x 10 Kw-hr of electricity, and in the 18 years of operation, no instability has been observed.
The staff finds that previous Dresden Ur.it No.1 cores have been designed to be
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9 aeutronically and thermal-hydraulically similar to previous fuel, and we find the above base's also acceptable to demonstrate the stability of the Cycle 11 core.
Further, the licensee has agreed to restrict operation of the Dresden Unit No.1 Cycle 11 core to preclude criticality unless at least two-recirculation pumps are ir. operation.
This restriction will provide a significant increase in reactor core stability mcrgins during Cycle 11 by preventing operation at any measurable power level while in the natural recirculation mode, thus preventing operation in the least stable mode for BWR's. This restriction will also preclude the possibility of a recirculation pump startup from natural circulation conditions, which could cause a reactivity insertion transient in excess of the most severe coolant flow increase currently analyzed.
We find the restriction acceptable for both purposes stated above.
2.7 Startup Tests CECO has proposed a series of startup tests (Reference 2a) and has 1
agreed to provide a report of those tests following startup.
We find the tests and reports proposed by CECO to be acceptable for the purpose of confirming predicted behavior of the Cycle 11 core.
3.0 TECHNICAL SPECIFICATION CHANGES We find the proposed Technical Specification changes acceptable and consistent with the information submitted'in support of the reload for Cycle 11. When the plant is operated in conformance with the proposed Technical Specification changes, it will be within the range of operating conditions assumed in the above described transients and accidents, which were found acceptable, and therefore such plant operation is acceptable.
- 4. 0 ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not involve a change in feffluent types or total amounts nor an increase in power level and will not. result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
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5.0 C0t:CLUSION
'n'e nave concluded, based on the considerations discussed above, that:
there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and that such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be _ inimical to the common defense and security or to the health and safety of the public.
Date:
September 20, 1977 I
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REFERENCES 1.
Letter to NRC from R. L. Bolger, Commonwealth Edison, Dresden Unit No.1 Cycle 11 Licensing Report, April 11, 1977.
2.
Letter to E. Case, NRC, from R. L. Bolger, Commonwealth Eoison, August 3,1977, with attachments:
- a. - Answers to NRC Round One Questions
- b. - Proposed Technical Specifications and Limiting Conditions of Operation
- c. - The XN-1 Heatup Analysis Supplement to Dresden Unit 1 LOCA Analysis 3.
Letter to NRC from R. L. Eolger, Commonwealth Edison, Answers to NRC Round Two Questions and Proposed Technical Specifications, September 8,1977.
4.
Memorandum and Order, Nuclear Regulatory Commission, August 21, 1975.
5.
Amendment No. 3 to NED0-20566, Effect of Steam Environment on BWR Core Distribution, General Electric Company, April 1,1977.
6.
hUS-3005, Big Rock Point Core Spray Test Report, Single Nozzle Test and Development Program, August,1977.
7.
Letter to Commonwealth Edison from NRC staff (K. Goller) with Round Two Questions, September 2,1977.
8.
Commonwealth Edison (M. Turbak) letter to NRC (D. Ziemann), Core Spray Distribution and Cooling Coefficient, April 2,1977.
9.
Commonwealth Edison letter to NRC, ATWS, October 1,1977.
10.
' etter to NRC from Commonwealth Edison, Technical Specification and Surveillance Requirements on Operating Limit MCHFR, September 15, 1977.
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- 11. High Energy Pipe Breaks Outside of Containment a.
J. S. Abel letter to A. Giambusso dated August 2,1973, transmitting:
" Criteria for Evaluation of Effects of Steam and Feedwater. Piping Breaks Outside Containment for Dresden 1" by NSC.
b.
J. S. Abel letter to A. Giambusso dated March 5,1974, transmitting:
Volumes 1 through 4 - Dresden Station Special Report No. 38 -
Evaluation of Steam and Feedwater Pipe Breaks Outside Containment Dresden 1.
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REFERENCES (Cont.)
.12.
Control ~ Rod Drop Accident a.
Test Design and Analysis Bulletin DNPS-1 XXVI-Reevaluation of Potential Reactivity Addition Accidents for-Dresden Unit No.1, July 29,1967.
- b. - J. S. Abel letter to D. J. Skovholt dated August 3,1973, Dresden Station Special ~ Report No. 32 - Reactivity Worths of Control Rods in the Dresden 1 Reactor.
13.
Refueling Accident M. S. Turbak letter to D. L. Zieman i, dated March 18, 1977, transmitting Evaluation of Refueling Accidents Dresden 1, 2 and 3 and Quad Cities 1 and 2.
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