ML17331A959
| ML17331A959 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 02/16/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17331A957 | List: |
| References | |
| NUDOCS 9309020112 | |
| Download: ML17331A959 (80) | |
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UNITED STATES NUCLEAR REGULATORYCOMMISSION WASHINGTON, D. C. 20666 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.
18 TO LICENSE NO. DPR-58 INDIANA & MICHIGAN ELECTRIC COMPANY INDIANA & MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C.
COOK NUCLEAR PLANT UNIT NO.
1 INTRODUCTION By letters dated July 20, and December 7, 1976, and February 4 and 9,
- 1977, supplemented by letters dated July 19, October 1, November 5, 17, 23 and 30, and December 7, 9 and 13,
- 1976, and February 8 and 9, 1977, Indiana 8 Michigan Electric Company and Indiana 8 Michigan Power Company (the licensee)'equested amendments (hereinafter referred to as amendment) to Facility Operating License No. DPR-58 for the D. C.
Cook Nuclear Plant Unit No.
1 (the facility).
The amendment would authorize operation with reactor power levels not in excess of 3250 megawatts (thermal) for core cycle 2 with (1) 65 Exxon Nuclear Company reload fuel assemblies, (2) an Exxon Nuclear Company emergency core cooling system
- analysis, (3) revised technical specification requirements for the ice condenser
- system, and (4) modifications to certain electrically operated valves to preclude single failures that would result in loss of emergency core cooling system capacity and to eliminate the need for actuation of the valves by personnel outside the control room.
The amendment also (5) corrects minor errors and inconsistencies in the technical specifi-cation requirements for containment air recirculation fan response
- time, containment penetration and valve leakage rates, the audit responsibili-ties of the Nuclear Safety and Design Review Committee, and safety related hydraulic snubbers.
BACKGROUND Operation of D.
C.
Cook Unit No.
1 at steady state reactor core power levels not in excess of 3250 megawatts thermal (100 percent of rated power) was authorized by Amendment No.
14 to Facility Operating License DPR-58 issued by the Commission on May 28, 1976.
This authorization is effective only until the reactor is shutdown for refueling at which time, unless the Commission takes further licensing action, the authorized power level would be 2632.5 megawatts thermal (81 percent of rated power).
This restriction on maximum power level after the first cycle, was made by the Nuclear Regulatory Commission in accordance with the advice of the Advisory Committee on Reactor Safeguards (ACRS) in the March ll, 1976 letter from Dade W. Moeller, ACRS Chairman, to the Honorable William A. Anders, Chairman of the Nuclear Regulatory Commission.
9309020112 770216 PDR ADQCK 05000315 P
For the first refueling of the D. C.
Cook Unit No.
1 reactor (scheduled for December 1976 - January 1977), the licensee has proposed to replace 65 of the original Westinghouse Electric Corporation fuel assemblies with Exxon Nuclear Company (ENC) assemblies and to demonstrate conformance of the facility's ECCS with the requirements of 10 CFR Part 50.46 by using an ENC ECCS analysis.
The Core Reload Evaluation section of this report addresses the licensee's proposal.
In the ACRS letter dated October 17, 1973, regarding D. C.
Cook and in the Regulatory staff evaluation of "Tests Conducted to Demonstrate the Functional Adequacy of the Ice Condenser Design" dated April 25, 1974, the need for a program to periodically measure the weights of selected ice baskets in the ice condenser was recognized.
In support of this program, the licensee has submitted, for our review, the results of the ice weighings since January 1976.
We have combined this data with previous weighing results and our evaluation is given in the section of this report entitled Ice Condenser Evaluation.
I In Section 5.4 of Supplement No.
5 to the D. C.
Cook Unit No.
1 Safety Evaluation Report, dated January
- 1976, we identified certain valves whose spurious actuation could adversely affect the performance of the ECCS following a postulated loss of coolant accident (LOCA).
The staff concluded that removing AC power from the valves would be an acceptable method to prevent such spurious actuation.
- However, several of the valves must be repositioned about 25 minutes after the postulated LOCA when the reactor cooling mode is shifted from injection to recirculation.
To operate these valves after the
- LOCA, operator action would be required to restore electrical power to the valves at the motor control centers outside of the control room.
The licensee committed to modify the control circuits of these valves. to eliminate the need for operator action outside the control room and preclude single failures that would result in spurious valve operation.
By letters dated February 27 and November 23,
- 1976, the licensee submitted proposed control circuit modifications for our review.
The acceptability of these modifications is discussed in this report.
C l
CORE RELOAD EVALUATION Discussion By Reference 1, Indiana and Michigan Electric Company and Indiana and Michigan Power Company (IBM) requested that the operating license of the Donald C.
Cook Nuclear Plant Unit No.
1 (License No. DPR-58) be amended to permit continued operation at steady state core power levels up to 3250 MWt (100% power).
By Reference 2, the licensee proposed changes to the Technical Specifications based upon an Exxon Nuclear Company (ENC)
ECCS evalua-tion model which conforms to the requirements of the Commission's regulations in 10 CFR 50.46.
The D. C.
Cook Unit No.
1 core consists of 193 fuel assemblies, each having a 15x15 array of fuel rods.
Each fuel assembly contains 204 fuel rods, 20 rod control cluster (RCC) guide tubes, and one instrumentatio'n tube.
Cycle 1 fuel was designed and fabricated by Westinghouse Electric Company.
For Cycle 2, 65 original fuel assemblies will be replaced by fuel assemblies which were designed and fabricated by the Exxon Nuclear Company (ENC).
The Westinghouse fuel remaining in the core during Cycle 2 (64 assemblies with an enrichment of 2.8'X U-235, and 64 assemblies with an enrichment of 3.3% U-235) will be scatter loaded throughout the interior of the core.
One Exxon fuel assembly will be loaded in the center of the core, and the remaining 64 Exxon fuel assemblies will be loaded in the core periphery.
The Exxon fuel has an enrichment of 2.95%
U-235; 16 of the new fuel assemblies have burnable poison rods-B4C pellets (8 assemblies having 8 burnable poison rods, and 8 assemblies having 4 burnable poison rods).
Mechanical Desi n
The Cycle 2 core will consist of 65 ENC assemblies and 128 Westinghouse assemblies.
The fuel assembly design parameters ay e shown in Table 1.
The ENC reload fuel is clad with Zircaloy-4 and prepressurized with helium.
One significant difference between the ENC and Westinghouse fuel is its clad thickness.
The ENC fuel cladding is 23'X thicker than the Westinghouse fuel cladding.
The ENC fuel also has shorter pellets than the Westinghouse fuel pellets.
The staff believes that
TABLE 1 FUEL ASSEMBLY DESIGN PARAMETERS*
Westinghouse Low Enrichment Fuel Westinghouse High Enrichment Fuel Exxon Fuel Enrichment (wt 'X U-235)
Number of Assemblies Pellet Density,
(%)
Pellet-to-Cl ad Gap (mil s)
Pellet Diameter (inches)
Fuel Stack Height (inches)
Number of Fuel Rods/Assembly Region Average Burnup at BOC2, (MWD/T)
Cladding Material Cladding OD (inches)
Cladding Thickness (inches)
Instrument Tube Material Instrument Tube OD (inches)
Spacer Grid Material Number of Spacer Grids 2.80 64 7.5
- 0. 3659 143. 4 204 18,100 Zircaloy-4 0.422 0.0243 Zircaloy-4 0.546 Inconel
- 3. 30 64 95 7.5
- 0. 3659 142.8 204 13,900
- 2. 95 65 7.5
- 0. 3565 144 204 0
Zircal oy-4 Zircal oy-4
- 0. 422
- 0. 0243
- 0. 424
- 0. 030
- 0. 546 Ineo nel
- 0. 544 Zircaloy-4 structural members with Inconel springs Zircal oy-4 Zircal oy-4
- initial unirradiated conditions
these design differences,-thicker cladding and shorter fuel pellets, are improvements with regard to pellet-cladding-interaction (PCI)
(Reference 27).
Hence the ENC fuel is expected to be more resistant to PCI than the original fuel.
The total weight of the ENC fuel bundles and the Westinghouse high enrichment fuel bundles does not differ by more than 2X.
The ENC fuel design for D. C.
Cook Unit 1 Cycle 2 is similar to that supplied by ENC for other facilities.
The cladding material, Zircaloy-4, was used in previous ENC fuel supplied for Palisades Core II, Yankee-Rowe Core XII, and H. B. Robinson Core IV.
136 assemblies were loaded into Palisades Core II, 40 assemblies were loaded into Yankee Rowe XII, and 52 assemblies were loaded into H.
B. Robinson Core IV.
The enrichment of the fuel for D. C.
Cook is in the range of that used in the above cores.
The general dimensions of the fuel rod
( including diametral gap which is of importance for stored energy) are within the range of PWR fuel designs previously irradiated successfully.
In response to our question regarding compatibility between the D.
C.
Cook Unit 1 fuel handling equipment, and the Exxon reload fuel, the licensee performed fit-up tests at the D. C.
Cook plant (Reference 4).
These tests indicated that there should be no diffi-culties in handling the Exxon fuel at the D. C.
Cook Nuclear Plant.
Approximately 1000 bundles manufactured by ENC are in-core, in PWRs and BWRs, with bur nups ranging from first cycle to 25,000 MWD/MTU.
Approximately 10% of these have exposures between 15-20,000 MWD/MTU.
Based on sipping results and surveillance of representative assemblies, no failures have been observed or detected.
The design of the ENC 15 x 15 reload fuel assemblies is described in Reference 3 which is a generic report giving a detailed description of fuel assembly design methods and bases.
Portions of thi s report regarding the effects of fuel densification have been reviewed by the NRC staff and found acceptable.
Other sections of the report are currently under review on a generic basis; and, therefore, have not been considered in our review of the use of ENC fuel in D.
C.
Cook Unit No. l.
Our conclusions concerning the acceptability of
the use of ENC fuel in D. C.
Cook Unit No.
1 are based on (1) the fuel design and analytical methods which have previously been reviewed by the staff, (2) the similarity of the reload fuel to that used in Cycle 1 which was previously found acceptable, and,(3) the successful operating performance of ENC fuel.
Based on these factors, we conclude that there is reasonable assurance that the performance of the ENC reload fuel will be acceptable.
In Reference 1, the licensee indicated that one or two fingers had broken off a control rod during drop timing tests which were performed during an April, 1976 outage.
The finger(s) which had detached from the rod control cluster assembly ar e presently in the fully inserted position.
Subsequent to the control rod finger failure the licensee performed analyses of the core to correlate the incore flux measurements with the known failures.
The licensee's study and measurements showed a slight skew in the bur nup of the fuel assemblies surrounding the the failed rod.
The licensee concluded that the effect is not signifi-cant enough to have a restrictive impact on the shuffling scheme for Cycle 2.
II In Reference 4, the licensee stated that he has reviewed this control rod problem with Westinghouse.
Based upon Westinghouse's extensive testing and evaluation program that was conducted prior to commercial use of these control rods, and based upon over 2700 rod-years of operation in commercial nuclear power plants, Westinghouse and the licensee believe that the control rod failure is an isolated incident not indicative of generic failures.
Westinghouse has examined whether control rod scram capabilities would be compromised by failure of additional rodlet/fingers.
After considering the possible rod failure modes Westinghouse concluded that such failures would not affect the reactor scram times.
As indicated in Reference 5, the licensee, by letter dated February 8, 1977, provided additional information concerning the failed control rod.
At the end of Cycle 1, rod drop timing and drag tests were performed on each control rod and all test results were within acceptable limits.
The results of the licensee's visual examinations of the affected control rod including the rod drive rod, guide tube removable insert, and the inside of the upper guide tube were provided.
The examination revealed that a two-rodlet vane had separated from the control rod hub at the vane-hub interface (at a tack welded and brazed joint); however, no cause for the failure could be determined.
The fuel assembly con-taining the broken rod has been removed from the core.
The 'licensee is continuing to investigate the cause of the failure and will provide additional data 60 days after startup.
1 Based on the above information, we have concluded that, although the specific cause of the failure has not been established, there is no evidence to indicate that this is a generic problem and, therefore, Cycle 2 operation of the D. C.
Cook Unit No.
1 need not be delayed.
We will, however, evaluate future information provided by the licensee on this subject.
Nuclear Desi n and Technical S ecification Chan es
~ ~ 3 Technical Specification changes required as a result of the nuclear design for D.
C.
Cook Cycle 2 are discussed in the following sections.
Shutdown Mar in In the analysis of the steam line break accident in Reference 6 for end of cycle 2 with the reactor at no load operating temperature, a
minimum shutdown margin of 1.75K hk/k is initially required to control the reactivity transient.
The corresponding shutdown margin for Cycle 1
was 1.6% gk/k.
Accordingly, the licensee has proposed to change the Technical Specification end of cycle shutdown margin requirement to 1.75'X sk/k.
We find this acceptable because it will prevent return to criticality in the event of a small steam line break and also provide an acceptable margin to DNB in the unlikely event of a large steam line break accident.
Power Distribution Control and Monitoring The ECCS analysis, Reference 7, was performed with an assumed heat flux hot channel factor, F~(Z) of 1.95.
The maximum F (Z) at full power for Cycle 1
was 1.98.
By letter dated Febr(ary 9, 1977, the licensee reported the results of an analysis of the effect of burnup on F (Z) using the NRC staff assumptions in Reference 15 for flow blockage cal-culations.
The licensee determined that the value of 1.95 would hold until a bur nup of 8500 MWD/MTU in Cycle 2 and then would decrease linearly to 1.90 at an expected end-of-life burnup of 10,800 MWD/MTU.
The D. C.
Cook Unit No.
1 Technical Specifications will be changed to reflect this behavior of the F~(Z) limit.
We find this to be acceptable.
The licensee will continue to use the Axial Power Distribution Monitoring System, APDMS, during cycle 2 to ensure that F (Z)
X P
(P
= fraction of full power) does not exceed the F(Z) likitduring normal operation.
The APDMS essentially performs direct measurements of the core peaking factor with in-core movable detectors and requires, by Technical Specification, power reduction and other appropriate actions if the peaking factor exceeds its limit.
Experience with the APDMS during Cycle 1 operation in D.
C.
- Cook, and in other reactors employing an APDMS indicates this system provides an adequate
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Data accumulated thus far support at least a
95% pro(ability with a 95% confidence level that the F (Z) will not exceed the APDMS measured value with the uncer taintie3 assigned to the APDMS and failure probabilities taken into account.
The power level, above which APDMS monitoring is required, is determined by the F (Z) which can be justified by monitoring with ex-core detectors.
For the latter part of Cycle 1
operation, the APDMS was required to be in operation above 90%
of full rated power.
This was a result of a plant specific analysis of ex-core detector monitoring using constant axial offset control (CAOC
) which indicated an F (Z) of 2.18 would not be exceeded using these procedures.
ThIIs 90%
< 1.98/2.18 X 100.
for Cycle 2 operation, Exxon has provided, in Reference 8,
an analysis of CAOC procedures which they term power distribution control (PDC
) which indicates F (Z) will not exceed 2.30 during normal operation.
Reference 8
fleas been reviewed and approved by the staff in Reference 23.
Based on the methods of Reference 8,
a new APDMS monitoring threshold of 1.95/2.30 x 100
= 84% is required from 0-8500 MWD/MTU with a linear decrease to 1.90/2.30 x 100
= 82% at 10,800 MMD/MWU to reflect the burnup dependence of F (Z).
The technical specifications will be modified to ref)ect these requirements.
At power levels up to the APDMS monitoring threshold, PDC proce-dures will ensure that the F (Z) limit assumed for the LOCA analysis will be maintained.
Above th(s level, the APDMS will provide the same assurance.
The PDC procedures are also required to be observed above the monitoring threshold to'nsure that axial power shapes not allowed by PDC do not occur and thus potentially violate DNB analyses.
This is presently required by the Technical Specifications.
We=-conclude the above provisions will -adequately ensure that initial conditions assumed for the LOCA and DNB analyses will be maintained during normal operation at power levels up to and including 100% rated power during Cycle 2.
The PDC study in Reference 8 addressed target offsets in the range
-7.5% to 0.0'X.
To increase plant operating, flexibilityusing
- PDC, ENC, by letter dated Feburary ll, 1977, submitted an addendum to Reference 8 which provided additional analysis with regard to positive target offsets.
Based on this addendum, we have determined that target offsets up to +5% at beginning of Cycle 2 decreasing linearly to +2%
for burnup of 7500 MWD/MTU and greater will continue to protect the PDC F limit and are, therefore, acceptable for Cycle 2 operation of D.
C.
Cook Unit No. 1.
These target axial offset values will be incorporated into the facility Technical Specifications.
Ph sics Test Pro ram The physics start-up test program proposed for Cycle 2
(Reference
- 5) is acceptable if the following guidelines are used in verifying predicted control rod bank reactivity worths and the shutdown margin.
Control rod bank worths must be measured for banks D, C, B, and A, individually. If any one bank worth differs from the predicted value by more than 15%, or the sum of the worths of the four banks differs from the predicted value by more than 10'X, the first shutdown bank should be measured.
If the sum of the worths of the control banks and the shutdown bank differs from the predicted value by more than 10%, additional shutdown bank measurements should be performed to verify technical specification shutdown margin.
The licensee will be required to include this test in the startup test program.
Anal tical Methods The analytical methods used by the licensee and ENC in the calculation of operational parameters for Core II are described in References 9 and 10.
These documents present the ENC neutronic design calculational methods along with the results obtained when these methods are compared to experimental measurements.
We have reviewed and approved these documents.
Therefore, we conclude that the analytical methods used to calculate the operational parameters for D. C.
Cook Unit No.
1 Core II are acceptable.
Thermal and H draulic Desi n
The thermal-hydraulic analyses of the Cycle 2 core (Reference 1) shows the following:
a.
The ENC and Westinghouse assemblies are thermally and hydraulically compatible.
b.
The minimum departure from nucleate boiling ratios (MDNBR) for both fuel types are always greater than 1.30 for normal operation and anticipated transients.
The analyses include both experimental measurements and theoretical calculations.
ENC has performed hydraulic flow tests to evaluate the compatibility between the ENC and the Westinghouse 15xl5 fuel assemblies.
The results of these tests show that even though the Westinghouse and ENC fuel assemblies exhibited some differences in the plenum-to-plenum pressure drops and the pressure drops between the tie plates, the difference in flow through the ENC and Westinghouse assemblies is small (average flow difference of 1.4% between the two types of fuel).
This difference of coolant flow has been considered in the analyses.
It has been determined that it has a negligible effect upon the margin to DNB.
The adequacy of the ENC fuel for meeting NDNBR requirements has been verified with transient analyses performed at 102'X power.
The results of the transient calculations are discussed later in this evaluation.
DNB calculations show that the RDNBR is greater than the minimum acceptable limit of 1.30 for both ENC and Westinghouse fuel assemblies under the operating conditions of Cycle 2.
Additional margin is provided by the fact that the steady state DNB calculations were performed at a stretch power level of 3640 tlWt although D. C.
Cook Unit 1 will be licensed for only 3250 NWt for Cycle 2.
Based on the above, we have concluded that the thermal and hydraulic design of the Cycle 2 core is acceptable.
On August 9, 1976 Westinghouse Electric 'Corporation presented data to the staff from recent experiments which showed that fuel rod bowing could have a significant effect on the departure from nucleate boiling ratio (DNBR).
In particular, these experiments showed that if a heated fuel rod was bowed to contact with an unheated rod (thimble rod),
a reduction in DNBR significantly greater than that expected would occur.
The staff has developed a model based on this data to calculate the DNBR reduction to be expected in operating reactors.
This model consists of three components.
- First, a method of calculating the clearance reduction between adjacent rods due to rod bowing is used to estimate the extent of fuel rod bowing for a given burnup.
- Second, using the 'Westinghouse data for DNBR reduction, the DNBR reduction for the calculated extent of rod bow is determined.
Finally, the calculated DNBR reduction may be offset by available margin.
D.
C.
Cook has margin available to offset the calculated reduction in DHBR, as discussed below.
For Cycle 2, D.
C.
Cook will operate with a combination of ENC and Westinghouse fuel.
ENC has presented no data on the extent of fuel rod bowing in ENC fuel; however, an analytical method of predicting fuel rod bowing has been presented to the staff (Reference 11).
This analytical method has not been accepted because we do not believe a mechanism for fuel rod bowing has been satisfactorily identified.
Thus a mechanistic calculation should not be employed.
We ther efore have
- assumed, as an interim position, that the amount of fuel rod bowing expected for ENC fuel will be equal to that expected for Westinghouse fuel.
This assumption is considered conservative because the thicker fuel rod cladding and slightly larger rod diameter of the ENC fuel provide a larger moment of inertia to resist bowing forces.
Since the Westinghouse and ENC fuel are of similar design, as described in the aiechanical
~Desi n section, the Westinghouse calculational model was used to determine the DNBR reduction for both fuel types (See Attachment A).
The maximum calculated reduction in DNBR for D. C.
Cook is 27.6'X.
The NRC staff has permitted licensees to offset calculated DNBR reductions by accounting for certain parameters which affect DNBR calculations for their plants (Attachment A).
For D.
C.
Cook Unit No. 1, the licensee has utilized the minimum DNBR of 2.01 which was calculated for the most limiting anticipated transient.
The difference between 2.01 and the current DNBR safety limit of 1.3 results in a 54.6X credit which more than offsets the DNBR reduction of 27.6'X.
Therefore, no changes in Technical Specifications are necessary to offset the effects of fuel rod bowing on DNBR during Cycle 2.
Transient and Accident Anal ses The licensee provided results of their ECCS analysis in References 2, 7, and 12, and descriptions and results of other transient analyses in References 2,
6 and 13.
ECCS Coolin Performance (LOCA) Anal sis Evaluation Model The licensee provided -Exxon's analysis of the ECCS cooling system performance.
(References 2, 7, and 12).
The model (Reference 14 )
addressed hot channel performance for the reload fuel, and the overall reactor response to the composite fueled core.
The calculational model used by Exxon for D. C.
Cook Unit 1 Cycle 2 was reviewed by the staff, and approved by the staff's Safety Evaluation (Reference 15 ).
The NRC staff, in Reference 15, specified assumptions to be used to determine the effects of fuel rod internal pressure on flow blockage calculations to demonstrate conformance to 10 CFR Part 50 Appendix K
criteria.
Using these assumptions, the licensee, by letter dated February 9,
1977, reported that the value of the F (Z) limit would be 1.95 until a Cycle 2 burnup of 8500 NWD/MTU at khich time the value would decrease linearly to 1.90 at 10,800 NWD/MTU.
The decrease in F (Z) is required to compensate for the assumed increase in fuel pqpn internal pressure as a function of burnup.
1r
~kS The worst break location was identified as the cold leg at the pump discharge.
For the first cycle, Westinghouse's analysis identified a double-ended guillotine break of the pump discharge line as the worst break.
ENC performed a series of break size calculations at the pump discharge line, assuming the worst single failure (loss of a low pressure ECCS pump, Reference 16).
The calculations were performed for double ended guillotine breaks with discharge coefficients of 1.0, 0.8, and 0.6.
The split break configurations were calculated with t)e break area equal to twice the cross sectional pipe area (8.25 ft ) and also for the2cases of the2flow area reduced to 0.8 and 0.6 times that area (6.6 ft and 4.95 ft respectively).
As shown in References 2 and 7, it was determined that the 8.25 ft split break is most limiting.
The maximum peak clad temperature was shown to be 2196'F, which is below the acceptable upper limit of 2200'F as specified in 10 CFR 50.46(b).
In addition, the maximum local metal/water reaction of less than 8% and the total core metal/water reaction of less than 0.8% were within the allowable limits of 17% and 1%
respectively.
These calculations were done using a total peaking factor of 1.95.
Sased on this analysis and the analysis of the effect of fuel pin internal pressure on F~(Z), the peak linear heat generation rate for the ENC fuel for Cycle 2 is 13.41 kw/ft until 8500 NWD/l1TU and then decreases linearly to a value of 13.06 kw/ft at 10,800 NWD/MTU, end-of-life.
With regard to small breaks, in Reference 2 the licensee indicated that the small breaks would result in conditions substantially below the limiting large break results and clearly within the requirements of 10 CFR 50.46.
We have reviewed the above results and agree that the break spectrum has been defined sufficiently to assure that the worst break size and location for D.
C.
Cook Unit 1/Cycle 2 has been identified and analyzed.
We find the break spectrum calculations acceptable.
Therefore, we have concluded that operation with the reload core consisting of Westinghouse and ENC fuel assemblies meets the requirements of 10 CFR 50.46 and is acceptable.
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C Post LOCA Lon Term Coolin Aa."<<gC>g, In Reference 2, the licensee informed the staff that the existing analyses which demonstrated the Emergency Core Cooling System's capability to meet the long term cooling requirements for Cycle 1
operations are valid and applicable for Cycle 2 operation.
We find this to be acceptable.
U er Head Tem erature Anal sis for ENC Fuel
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In Reference 12, ENC reported the results of studies performed to determine the sensitivity of LOCA calculations to upper head temperature.
These studies verified that the use of hot leg temperature for the upper head is conservative for ENC fuel.
Consequently, the ENC full,break spectrum analysis which was performed with the hot leg temperature assumed for the upper head is conservative and the results are acceptable.
U er Head Tem erature Anal sis for the Remainin Westin house Fuel The licensee has submitted a reevaluation of ECCS performance for D. C.
Cook Unit No.
1 (Reference
- 24) in response to our Order for Modification of License issued on August 27, 1976.
The reevaluation was made using the October,.1975 version of the Westinghouse ECCS Evaluation Model assuming the upper head fluid temperature equal to the fluid outlet (hot leg) temperature.
This analysis supersedes the. previously performed ECCS evaluation which used the same
- October, 1975 version of the evaluation model but which was based on the assumption that the upper head temperature was equal to the
cold leg temperature.
The reevaluation of the ECCS performance in Westinghouse plants was required because recent experimental data had indicated that the actual temperature in the upper reactor vessel head was in the range of 50-75 percent of the difference between vessel inlet and outlet temperatures (Reference 25).
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The reanalysis consisted of the evaluation of ECCS performance for double ended cold leg guillotine breaks (DECLG) with a discharge coefficient C
of 0.8.
The licensee claimed that this break size was reprIIsentative of the limiting value of peak clad temperature and Zr-H20 reaction.
To justify limiting the ECCS analysis to only one break size, the licensee referenced the previously approved ECCS analysis and the Westinghouse topical report WCAP-8855 which provided sensitivity studies for four loop
( 15 x 15 ) plants and which also had been reviewed and approved by the staff (Reference 26).
The previous ECCS analysis was performed for a spectrum of four breaks specific for D. C.
Cook using the October, 1975 evaluation model with the assumption of upper head temperature equal to the cold leg temperature.
This analysis identified the worst break size as the DECLG with C
= 0.8.
In addition, the sensitivity studies performed by Westinghouse and reported in Reference 25 indicated that, for a specific plant, the change of upper head temperature from cold to hot leg temperature did not affect the critical break type or size.
Based on these references, the licensee has concluded that the break size analyzed is the critical break for the D. C.
Cook Plant resulting in the peak clad temperature of 2164'F and the maximum local Zr-H20 reaction of 6.39 percent.
Based on our review of the submitted documents, we conclude that the results of the ECCS reanalysis, using the October, 1975 version of the Westinghouse ECCS Evaluation Model with upper head tempera-ture equal to the outlet (hot leg) fluid temperature, are con-servative relative to the 10 CFR 50.46 criteria, and are acceptable.
CONTAINMENT LOCA ANALYSES We have evaluated the effects of the D. C.
Cook Unit No. 1. Cycle 2 core on containment pressure response following a postulated LOCA.
Since the ENC reload core has been designed to the same thermal power rating (3250 megawatts) as the original core, only the core stored energy could alter the blowdown used for the original containment analysis.
The thicker clad of the ENC reload fuel results in an increase of 1.5% in the core stored energy.
Because core stored energy released to containment constitutes only about 2.5'X of the total energy released, the ENC fuel will result in approximately a
.04% increase in the integrated energy released to the containment at the time of ice melt.
This increase is negligible in comparison to the conservatisms in the currently approved Westinghouse containment analysis mass and energy release model for D. C.
Cook Unit No. 1.
Therefore, we conclude that there is a negligible change in the LOCA containment analysis as a result of the Cycle 2 core.
- 14 Conclusions Based upon the above information we have concluded that:
The ECCS cooling performance conforms to the peak clad temperature and maximum oxidation and hydrogen generation criteria of 10 CFR 50.46(b).
In addition the plant will conform with the criteria to maintain a eoolable geometry and provide satisfactory long term cooling.
The LOCA analyses assumed that there was a coincident loss of offsite power at the initiation of the accident, which would result in pump coastdown.
Since these analyses were performed for only four loop operation, and since other modes of operation have not been demonstrated to meet paragraph 50.46, reactor operation will not be permitted with one or more idle loops.
Rod Ejection Accident In Reference 2, the licensee provided an analysis for the rod ejection incident for fuel Cycle 2.
The licensee determined that for the worst case event the fuel limits would not be exceeded.
We find this acceptable.
Rod Dro Transient In Reference 2, the licensee provided information on analysis of the Rod Drop Incident for fuel Cycle 2.
The analysis showed that the results for the dropped rod incident and for the dropped bank incident for Cycle 2 are more favorable than those for Cycle l.
We find this acceptable.
Rod Withdrawal Transient The licensee has provided the results of a reanalysis of the rod withdrawal transient from full power using the Exxon PTSPWR2 Code (Reference 17 ).
We previously reviewed this code and found its results to be acceptable.
The rod withdrawal transient was analyzed from an initial power of 3315 NWt (102% power) for both slow and fast rod withdrawal as shown on Table 2.
The slow rod withdrawal results in the more severe conditions, but still within the Technical Specification limits (MDNBR = 2.15 vs. lower limit of 1.30, maximum pressurizer pressure of 2279 psi a vs.
2750 p'sia);
therefore we find these results to be acceptable.
TABLE 2 SUMNRY OF RESULTS FOR ENC FUEL Transient Initial Conditions For Transients Uncontrolled Rod Mithdrawal 88.0 x 10-4hp/sec Uncontrolled Rod Mithdrawal 8 2.0 x 10 4p/sec Loss of Flow-4 Pump Coastdown Loss of Flow-Locked Rotor Loss'f Load Large Steam Line Break Small Steam Line Break Maximum Power Level HWt 3315.
4230.
3633.
3315.
3315.
3321.
406.
Maximum Core Average Heat Flug Btu/hr-ft~
210,500.
222,100.
228,200.
210,500 2100500 210,500 23,140.
Maximum Pressurizer Pressure sia 2220.
2230.
2279.
2256.
2242.
2538.
MDNBR M-3 2.43 2.24 2.15 (1,93)t.
2,01 (1.86)+
1.98 2.43 2.90
- Pressure decreases from initial value.
'*.The core does not return to criticality.
+ Mith rod bow penalty
6 Loss of Coolant Flow Transient The analysis of the initial reference cycle showed the loss of coolant flow incidents, pump coastdown and locked rotor, to be the most limiting with respect to DNB (MDNBR's of 1.40 and 1.07 for the pump coastdown and the locked rotor respectively).
ENC 's reanalysis of these incidents resulted in NDNBR's of 2.01 and 1.98 for the pump coast-down and locked rotor cases respectively.
ENC's analysis shows that the maximum pressurizer pressure for these events was 2256 psia (for the pump coastdown).
'The l1DNBR's and maximum pressurizer pressures for these events are within the Technical Specification limits (DNBR > 1.30 and pressurizer pressure
< 2750 psia).
We find this to be acceptable.
Loss of Load Transient The loss of load transient was analyzed for the second cycle.
This transient was limiting with respect to system pressure.
For Cycle 2
the maximum pressurizer pressure calculated for this event was 2538 psia whereas the Technical Specification limit is 2750 psia.
The MDNBR and maximum pressurizer pressure are well within the Technical Specification limits and therefore are acceptable.
Other Transients and Accidents The kinetics parameters for the remaining transients and accidents are within the envelope of parameters analyzed for the reference cycle.
Therefore, the results of the reload cycle will be bounded by those for the reference cycle.
We find this to be acceptable.
~ P E
E ICE CONDENSER EYALUATION Since our January 1976 report on the status of the ice weight surveillance program of the D. C.
Cook Nuclear Plant Unit No.
1 presented in Supplement 5 to the Safety Evaluation Report (SER),
the licensee has performed four additional ice weighings.
The results of all ice basket weighing programs and the licensee's conclusions and recommendations have been documented in five reports (References 18 through 22).
Our evaluation and conclusions are based upon review of the information presented in the referenced documents.
In January, April, July, and September of 1976 the licensee weighed a sample population of ice baskets as a part of the continuing long-term evaluation of the ice condenser system.
The sample populations were composed of 166 baskets in January, 172 baskets in April, 177 baskets in July, and 179 baskets in September.
Analysis of these data and comparison with prior data indicate that the average ice loss rate continues to be about 2%/yr, (28 0/yr/basket) with a statistical maximum of about 2.5%/yr, (35 0/yr/basket).
Data from the March, July, and October 1975 basket weighings revealed that ice loss is not uniformly distributed over the ice condenser.
A pattern of preferential ice loss was evident, with the baskets in rows closest to the containment wall having the lowest ice loss and the losses becoming progressively greater as the basket positions approach the crane wall.
The additional data taken in April, July and September of 1976 have confirmed the existence of the preferential loss pattern within the ice condenser.
Analysis of the data for the period from April 1976 through September 1976 indicates that ice baskets adjacent to the containment wall (radial row 1) have an average ice loss rate of about 1/2%/yr while those adjacent to the crane wall (radial row 9) have ice loss rates averaging about 5-3/4%/yr.
In Supplement No.
5 to the
- SER, we reported that the licensee had developed special weighing equipment which permitted successful weighing of baskets in radial rows 1
and 9 (wall baskets adjacent to the containment wall and crane wall, respectively) during the July and
- October, 1975 weighing programs.
In April of 1976, the licensee weighed the wall baskets with an improved model of the wall basket weighing device.
The improvements to the wall basket weighing device have resulted in a signficant reduction in repeatability error associated with the weighing of wall baskets, such that the repeatability error is now comparable with the error associated with
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weighing the remainder of the ice baskets.
Analysis of the data taken with the improved wall basket weighing device now indicates that the average ice loss rate in radial row 9 (wall baskets adjacent to the crane wall) is greater than the average loss rate in radial row 8.
Average loss rates in radial rows 1
and 2 are significantly less than in rows 8 and 9.
Also, rows 1
and 2 do not exhibit a marked difference between the average ice loss rates.
The average loss rates for ice baskets in radial rows 1, 2, 8 and 9 for the period of April 1976 to September 1976 are 7 ¹/yr (1/2%/yr), 9.6 ¹/yr (3/4%/yr),
66 ¹/yr (4-1/2%/yr) and 82 ¹/yr (5-3/4%/yr) respectively as indicated in Table 3.
The distribution of ice within the ice condenser is shown in Figures 1
and 2 on an average weight per bay basis and an average weight per radial row basis, respectively.
These figures indicate the distribu-tion of the ice last measured in September
- 1976, and the projected distribution of the ice based on measured average loss rates and the uncertainty associated with the measurement of loss rates at a 95%
level of confidence.
The projected distributions, therefore, represent the minimum expected ice weights in the ice condenser for two different future times.
The selected points in time are February 1977 and May 1978.
We expect February 1977 to be the approximate time for the next ice condenser weighing program and the completion of the first reactor refueling and subsequent plant start-up for fuel cycle 2.
May 1978 represents the maximum expected life of fuel cycle 2 based on a
design fuel life of 12 months and a
25'X contingency.
As we repor ted in Supplement No.
5 to the SER, the minimum amount of ice uniformly distributed throughout the ice condenser to prevent containment overpressurization in the event of the design basis accident is 1098 pounds/basket.
With this as a basis, we have previously established a Technical Specification average weight limit of 1220 pounds/basket for initiation of an operating period (i.e.,
operability of the ice condenser).
This limit is established to assure that during the specified operating period the average weight of any significant group of ice baskets will not be less than the minimum uniformly distributed amount of ice (1100 pounds/basket) assumed in the design basis accident analysis.
As may be seen from Figures 1
and 2 using the latest measured average ice weights and the maximum projected loss rates at a 95'4 level of confidence we would not project any bay or radial row of baskets to fall below the 1100 pound/basket value used in the design basis accident analysis during Cycle 2.
4 r,'XI,,
i Continued weighings of the weighable basket nearest the ice condenser lower plenum personnel access door (basket 2-8 in bay 24) indicate a
local area of greatly increased loss rate.
The current rate of ice loss for basket 2-8 is about 155 pounds per year (12%/yr) and the measured weight of ice in the basket in September, 1976 was 1128 pounds.
Clearly basket 2-8 and the surrounding baskets (i.e., baskets 1-7, 1-8, and 1-9 and 2-7 and 2-9) in bay 24 would be expected to weigh less than the 1100 pounds assumed in the accident analysis before the start of cycle 2.
As a result, the licensee has committed to add approximately 200 pounds of ice to each of the six baskets indicated above during the current refueling outage, by a technique demonstrated during the April 1976 basket weighing program.
The licensee and Westinghouse have developed a method by which up to 300 pounds of ice may be added to an individual basket, by drilling a 2 inch diameter hole in the upper six feet of a basket and "trickling" a 34'F solution of bor ated water into the basket over an extended duration.
This method of ice addition appears practical when only a few baskets are involved, but has yet to be proven as a feasible method of ice addition if entire bays or radial rows would require an ice addition.
Continued development efforts by the licensee and NRC staff review of procedures and equipment are necessary to permit the large scale addition of ice to signficant groups of baskets.
The alternative to large scale ice addition is the complete melt-out of the ice condenser and refilling the ice baskets, a process which would require 3-6 months to complete.
In the October report (Reference
- 5) the licensee submitted the results of an analysis of the plant response to the design basis accident assuming a maldistribution of the ice in the ice condenser.
The analysis shows that the design pressure of the containment (12 psig) is not exceeded when the average ice weight in two bays
( 162 baskets) is 850 pounds per basket and the average ice weight in the remaining 22 bays
( 1782 baskets) is 1120 pounds/basket.
An analysis of this type which recognizes the measured distribution of the ice inventory may be required in the future to demonstrate the acceptability of the ice condenser for continued operation.
- However, we believe that operation of the ice condenser with known groups of baskets below an average weight of 1220 pounds/basket should not be permitted solely on the basis of the licensee's analysis un'til the staff has a confirmatory long term containment analysis capability.
We expect the CONTEMPT-4 long term containment code with ice condenser modeling will be available by September 1977.
After reviewing the five reports by the American Electric Power Service Corporation regarding the basket weight history and analysis during the first twenty (20) months of plant operations, we have reached the following conclusions regarding the future operation of D.
C. Cook, Unit No. 1, and analysis of the ice condenser:
l.
Sufficent data have been collected to conclude that the plant can be operated safely at the full design power level for the expected life of the second reactor core (i.e., until about May 1978).
2.
Calculations of average ice weight per basket for a bay, a radial row or the total ice condenser should be biased to account for lighter ice baskets under the intermediate deck center support beams identified during the December 1974 basket weighing program.
This conclusion was identified and the basis discussed in Supplement No.
5 to the SER.
3.
Based on our review of the rate of ice loss and the pattern of loss in the ice condenser, we expect that the ice condenser may not have sufficient ice inventory to allow initiation of operation of the plant for fuel cycle 3.
As a result, it appears that additional emphasis should be placed on the development of ice addition techniques and equipment.
Analysis of the con-tainment considering the measured and projected distribution of the ice may also be required.
It should be noted that the development of a confirmatory long term ice condenser contain-ment code for the staff will be required to confirm the continued safe operation of the plant with maldistribution of ice, without requiring a complete melt-out and refilling of the ice condenser.
It appears that these developments will be required before cycle 4 operation and could possibly be required prior to cycle 3 operation.
4.
Me have determined that the following changes to Technical Specification 3/4.6.5 regarding the minimum ice weight for operation of D.
C.
Cook Nuclear Plant, Unit No. 1, are required.
The changes would:
a.
increase the number of ice baskets to be weighed, b.
increase the ice basket weighing frequency, and c.
assure sufficient ice for continued operation on a radial row basis as well as a bay by bay basis.
p/
The licensee has agreed with these conclusions
- and, by letter dated December 7,
- 1976, has proposed changes to Technical Specification 3/4.6.5.
The proposed changes would (1) increase the minimum number of baskets to be weighed from 96 to 144 and -would include baskets from radial rows 1
and 9, (2 ) increase the inspection frequency from 18 months to 12 months, and (3) demonstrate a
sufficient ice inventory in specific groups of baskets on a radial row basis.
Ne have concluded that the proposed Technical Specifica-tions are consistent with the results of our evaluation.
The increase in the minimum number of baskets to be weighed will assure that sufficient data are obtained to continue to evaluate the pattern and extent of preferential sub-limation losses in the ice condenser and, to the maximum extent practical, at least one basket is weighed in each bay from a location where maximum ice loss is expected to occur.
The increased inspection frequency is consistent with the maximum expected life of fuel cycle 2 and the demonstration of sufficient ice inventory on a radial row basis will properly account for the observed preferential ice loss patterns.
Me conclude that the proposed ice weighing Technical Specifications are acceptable.
5.
Additional weighing of the ice baskets in accordance with Technical Specification 3/4.6. 5 will be required before the Technical Specification may be modified to include the measured ice loss rates of the ice condenser during normal plant operation.
As indicated in our bases for the Technical Specifications, we believe that data used to calculate a representative ice loss rate for the ice condenser during normal plant operation should be obtained over a period of at least three years.
- I V'W I.
4 v
In a letter dated December 7, 1976, the licensee proposed changes to the ice condenser technical specification.
These changes would (1) reduce the surveillance interval for measurement of ice condenser inlet door opening, closing, and frictional torques from 18 months to 6 months, (2) reduce the surveillance interval for verifying the intermediate deck doors were closed and free of frost accumulation from 3 months to 7 days, and (3 ) correct an inconsistency between Table 3.3-5 and Specification 4.6.5.6.a concerning the response time of the containment air recirculation fan.
The first and second proposed changes are consistent with the licensee's initial ice condenser operating experience and with observations made during a review by the Office of Inspection and Enforcement of the adequacy of the ice condenser Technical Specifications.
The following information concerning ice condenser inlet and intermediate deck doors was compiled by the staff:
The inlet doors to the ice condenser have been found to have higher than allowable opening torques on approximately 8 occasions when inspected each 90 days during the period November 1974 to present.
On each of these occasions, 1 to 2 doors have been found with higher than permissible opening torques because of seal freeze up.
The technical specification limit on opening torques is less than 675 inch pounds.
The inlet doors were typically found with opening torques in the range of 800 to 1200 inch pounds.
2.
The original analysis of the Cook facility indicated that up to 8 of the 48 inlet doors could fail to open during LOCA conditions with acceptable consequences.
3.
Mestinghouse Electric Corporation has installed several new prototype seals in several inlet doors to confirm their design suitability under actual operating conditions.
These seals presently have about.9 months of operating experience.
One design appears to show decreased freeze up tendencies although it also appears to have corner sealing problems which will require further modification.
The presumption is that when a suitable seal is developed, it will be installed in the D. C.
Cook facility and incorporated into future ice condenser plants.
4.
Inspection of the inlet doors has to be performed with the plant at zero power because of the high radiation levels in their vicinity during power operation.
Inspection of the intermediate deck doors can be performed during power operation without excessive personnel exposure.
5.
Although the Technical Specifications require inspection of the inlet and intermediate deck doors on an 18 month interval, the licensee is presently inspecting the inlet doors each 3 months and the intermediate decks doors each 7 days.
6.
The tendency for the inlet doors to exceed the specified torques is reduced by more frequent exercising of the doors.
7.
Intermediate deck doors may become inoperable because of ice formation from condensation.
Based on the above considerations, we have concluded that the proposed Technical Specifications changes for surveillance of inlet and intermediate deck doors will improve the operability of the ice condenser and are acceptable.
We also have determined that the proposed change to Table 3.3-5 does remove an inconsistency between that table and Specification 4. 6. 5. 6. a and is acceptable.
I CONTROL CIRCUIT MODIFICATIONS FOR CERTAIN ELECTRICALLY OPERATED VALVES By letter dated November 23, 1976, the licensee revised the proposed control circuit design for eight electrically operated valves which was previously submitted on February 27, 1976.
The modifications are designed to eliminate the need for operator action to restore power to the valves from outside the control room and precludes a single failure which could cause a loss of ECCS cooling capability.
The modifications to the valve control circuits consist of the addition of (1) a key-lock feature for the control switch, (2) separate control power lockout switch, (3) annunciation of control power not locked out, and (4) valve position indication when valve control power is deenergized.
The licensee has developed test procedures which will detect single electrical failures during periodic surveillance testing of the control circuits.
Although only five of the eight valves in question must be reposi tioned during the switch over from injection to recircu-lation cooling flow, the licensee decided to modify all the valves to eliminate the need for operator action outside the control room in the event any of the valves had to be operated.
We have reviewed the revised modifications described in the licensee's November 23, 1976 letter.
Based on this review, we have concluded that the modified design for remote actuation of the valves from the control room satisfies the requirements of the single failure criteria and is acceptable.
CONTAINMENT AIR RECIRCULATION FAN AND LEAKAGE RATE TECHNICAL SPECIFICATIONS In two letters dated December 7, 1976, the licensee requested changes to the Technical Specification requirements for con-tainment air recirculation fan response time and containment valve and penetration leak rates.
The requested changes would (1) alter the response time of the air recirculation fans in Technical Specification Table 3.3-5 from <600 seconds to
<660 seconds and would (2 ) remove statements which indicate that Table 3.6-1 provides a list of all valves and penetrations subject to Type 8 or C tests, as defined in 10 CFR Part 50, Appendix J.
We have determined that requested change
( 1) would eliminate an inconsistency between the requirements of Table 3.3-5 and the Surveillance Requirement of specification 4.6.5.6.a which lists the response time for containment air recirculation fans as 10+1 minutes.
This requested change is acceptable.
We have determined that Technical Specification Table 3. 6-1 is not intended to list all containment valves and penetrations subject to Type B or C tests.
Therefore the requested change to Table 3.6-1 is acceptable.
NUCLEAR SAFETY AND DESIGN REVIEW COMMITTEE (NSDRC) AUDIT RESPONSIBILITIES By letter dated February 4, 1977, the licensee requested changes to the Technical Specification requirements for the audit responsibilities of the NSDRC.
The change would remove the word "all" from the speci-fication which now requires the NSDRC to audit:
1.
The conformance of facility operation to "all" provisions contained within the Technical Specifications and applicable license condi-tions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
The results of "all" actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months.
3.
The performance of "all" activities required by the (}uality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.
We have reviewed the proposed Technical Specification change and have determined that it is in conformance with the current NRC requirements and with the Standard Technical Specifications which the NRC currently applies in the licensing of new facilities.
Therefore, the requested change is acceptable.
SAFETY RELATED HYDRAULIC SNUBBERS By letter dated February 9, 1977, the licensee proposed to alter the location description for snubber 811928 in Table 3.7-4 of the Technical Specifications.
The change in location description was found to be necessary because this snubber was not installed in the proper location.
The licensee has moved the snubber to the pr oper location and the requested Technical Specification is necessary to identify this new location.
We find the proposed technical specification acceptable because it results from the correction of an installation error for snubber 811928 and, therefore, corrects an error in the Technical Specifications.
TECHNICAL SPECIFICATIONS The proposed Technical Specification changes for D. C.
Cook Unit No.
1 Cycle 2 operation include:
l.
ENC 15x15 reload fuel limits.
2.
3.
Ice condenser surveil lance requirements.
4.
Specifications regarding control circuit modifications for certain electrically operated valves.
5.
Correction of errors for:
- containment air recirculation fan response time
- containment valve and penetration leak rates
- audit responsibility of the NSDRC
- safety related hydraulic snubbers Some modifications to the proposed Technical Specifications were necessary to meet NRC staff requirements.
We find the proposed Technical Specifi-
- cations, as modified, to be acceptable and consistent with the information submitted by the licensee.
s 1
y s REPORT OF THE ACRS At its 201st meeting on January 6-8, 1977, the Advisory Committee on Reactor Safeguards (ACRS) reviewed the proposal to refuel D. C. Cook Unit No.
1 with a par tial loading of ENC fuel assemblies and to subse-quently operate the facility at full rated power.
A copy of the Committee's report of its review dated January 14, 1977 is enclosed as Attachment B.
The ACRS had previously discussed D. C.
Cook Unit No.
1 in its reports dated December 13,
- 1968, October 17,
- 1973, and March 11, 1976.
The Committee concluded that full power operation of the proposed reload core was acceptable.
However, the Committee requested to be kept informed with regard to fuel pellet-clad interaction and fission gas release rate of the Westinghouse fuel for operation near the end of Cycle 2 and with regard to the control rod fingers which had broken during rod drop timing tests.
The staff will keep the Committee informed regarding fuel pellet-clad interaction and fission gas release rate as the results of future analyses become available.
The staff's conclusion with respect to the broken control
'od fingers has been discussed in a previous section of this evaluation.
The decrease in F from 1.95 to 1.90 as 'a function of burnup, which is described in tlIe ECCS LOCA Anal sis section of this evalualtion, was not discussed at the January 6-8, 197 ACRS meeting because this behavior of F was not known at that time.
The burnup dependence of F~ resulteII from the incorporation into the D. C.
Cook Unit No.
1 ECCS analysis of a revised flow blockage model.
The revised flow blockage model was set forth in the NRC staff's January 5,
1977 safety evaluation of the ENC ECCS model (Reference 15).
Although an F
value of 1.90 is relatively low and the Committee has expressed concer n over low peaking factors in the past, we believe, as stated in Reference 28, that the conservatisms (such as, the decay heat model) in the Appendix K criteria provide a high margin of safety and, since operation within F
limits as described in the ECCS LOCA Ana'I sis section fulfills the A pendix K requirements, no additional margin is required.
It bears repeating that the higher peaking factor (FO = 1.951 will be the limiting condition for about 80% of the fuel cycle.
Additional comments by committee members David Okrent and Milton Plesset are attached to the January 14, 1977 ACRS report.
These comments deal with the staff's treatment of ECCS evaluations.
These comments are all generic in nature in that they are applicable not only to the D.
C.
Cook Unit No. 1, but also to other licensed facilities.
The staff is con-sidering these opinions in this generic context, and will publish a
report to the ACRS on its conclusions thereon in the near future.
ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not authorize a change in effluent 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 or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
CONCLUSION Based on our review of the items identified as (1) and (2) in the introduction to this evaluation, and the considerations discussed in this evaluation, we have concluded that there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner.
Based on our review of the remaining items identified in the introduction to this evaluation of this evaluation and the considerations discussed in this evaluation, we have concluded that
( 1) because the items do not involve a significant incr ease in the probability or consequences of, accidents previously considered and do not involve a significant decrease in safety margin, they do not involve a significant hazards consideration, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner.
We also have concluded, based on the considerations discussed in this evaluation, that all of the activities discussed herein will be conducted in compliance with the Commission's regulations and the issuance of an amendment to the license will not be inimical to the common defense and security or to the health and safety of the public.
Date:
February 16, 1977 Attachment A:
Revision 1 to Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Calculations, dated February 16, 1977 :
ACRS Report dated January 14, 1977
REFERENCES l.
Indiana and Michigan Power Company letter (Tillinghast) to USNRC (Rusche) dated July 20, 1976, transmitting XN-76-25 "Donald C.
Cook Unit 1 cycle 2 Reload Fuel Licensing Data Submittal."
2.
Indiana and Michigan Power Company letter (Tillinghast) to USNRC (Rusche) dated November 23, 1976.
3.
"Generic Fuel Design for 15xl5 Reload Assemblies for Westinghouse Plants,"
XN-75-39, September, 1975.
4.
Indiana and Michigan Power Company letter (Tillinghast) to USNRC (Rusche) dated November 5, 1976.
5.
Indiana and Michigan Power Company letter (Tillinghast) to USNRC (Rusche) dated October 1,
1976.
6.
"Plant Transient Analysis for the Donald C.
Cook Unit 1 Nuclear Power Plant," XN-76-35, August, 1976.
7.
"Donald C.
Cook Unit 1
LOCA Analyses Using the ENC WREN-Based PWR ECCS Evaluation Model (ENC-HREM-II)" XN-76-51 October 1976.
8.
"Exxon Nuclear Power Distribution Control for PWR's," XN-76-40, September, 1976.
9.
"Exxon Nuclear Neutronic Design Methods for Pressurized Water Reactors,"
XN-75-27, June, 1975.
10.
"Exxon Nuclear Neutronic Design Methods for Pressurized Water Reactors,"
XN-75-27 Supplement 1, September, 1976.
ll.
"AXIBOW Code Manual for Rod Bowing Calculational Procedure,"
XN-CC-31, April 30, 1975.
12.
Exxon Nuclear Company, Inc. letter (Owsley) to USNRC (Ziemann) dated November 30, 1976.
13.
"Assumptions Used in the Plant Transient Analysis for the Donald C.
Cook Unit 1 Nuclear Power Plant," XN-76-35 Supplement 1,
- November, 1976.
g>
14.
15.
16.
17.
18.
19.
20.
21.
22.
"Exxon Nuclear Company WREN-Based Generic PWR ECCS Evaluation Model (ENC-WREM-II) 4-Loop PWR with Ice Condenser Large Break Example Problem," XN-76-36, August 1976.
"Revision 1 to the Staff Safety Evaluation Report on the Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC-WREM-II For Conformance to Requirements of Appendix K to 10 CFR 50," Office of Nuclear Reactor Regulation, January 5,
1977.
Indiana and Michigan Power Company letter (Tillinghast) to USNRC (Rusche) dated December 9, 1976.
Kahn, J. D., "Description of the Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTSPWR),"
XN-7774-5 Rev. 1, May, 1975.
"Long Term Evaluation of the Ice Condenser System - Results of the December 1974 Initial Ice Weighing Program," American Electric Power Service Corporation, J.
G. Feinstein, March 1975 (revised December 4, 1975).
"Long Term Evaluation of the Ice Condenser System - Results of the March 1975 Ice Weighing Program,"
American Electric Power Service Corporation, J.
G. Feinstein, June 1975.
"Long Term Evalua'tion of the Ice Condenser System - Ice Loss Calculations Resulting from the July 1975 and October 1975 Ice Weighing Programs,"
American Electric Power Service Corporation, J.
G. Feinstein, November 1975.
"Long Term Evaluation of the Ice Condenser System - Results of the January 1976 and April 1976 Ice Weighing Programs,"
American Electric Power Service Corporation, J.
G. Feinstein, and G.
M.
Gurican, July 1976.
"Long Term Evaluation of the Ice Condenser System - Result of the July 1976 and September 1976 Ice Weighing Programs,"
American Electric Power Service Corporation, J.
G. Feinstein, G.
M. Gurican, and J.
G. DelPercio, October 1, 1976.
23.
"Safety Evaluation Report on Exxon Nuclear Power Distribution Control for Pressurized Water Reactors",
XN-76-40, Office of Nuclear Reactor Regulation, February 8, 1977.
1 pg I
24.
Indiana 8 Michigan Power Company letter (J. Tillinghast) to USNRC (B. Rusche),
dated December 7, 1976, transmitting "P2B-ECCS Analysis of 0.8 DECLG with Upper Head at THOT".
25.
Letter NS-CE-1163, C. Eicheldinger (Westinghouse) to V. Stello (NRC) dated August 13, 1976.
26.
"USNRC Evaluation of Topical Reports on Effect of Upper Head Fluid Temperature on ECCS Performance",
December 17, 1976.
27.
"Studies of Fuel Clad Mechanical Interaction and the Resulting Failure Mechanism", E.Rolstad, K. D. Knudsen, Nuclear Technology Volume 13, February, 1972.
2S.
Supplement No.
6 to the Safety Evaluation Report Related to Operation of Donald C. Cook Nuclear Plant Unit No. 1, Office of Nuclear Reactor Regulation, May 1976.
ATTACHNENT A INTERIM SAFETY EVALUATION REPORT ON EFFECTS OF FUEL ROD BOWING ON THERMAL tQRGIN CALCULATIONS FOR LIGHT WATER REACTORS (REVISION I)
February 16, 1977
CONTEHTS 1.0 Introduction 2.0 DiNBR Reduction Due, to Rod Bow 3.0 Application To Plants In The Construction Permit And Operating Lice'nse Review Stage 4.0 5,0 Application To Operating Reactors References
] t I i.rndu(
i)il Data llave receiitly been preseiited (I<<f ren.".e 1) to the std 'f whi '.h show that previous'ly develo,>ed metliods for accounting For the eFfect.
of fu
'I rod boiving nii cl "Impart,i>> e fioi:i nuclMt>> boiling iii a pressurized water r eactor (PHR) may not contain adequate thermal margin when unheated
- rods, such as instrument tubes, are present.
Further experimental verification of these data is in progress.
However an interim measure is required pending a final decision on the validity of these new data.
The staff has evaluated the impact of these data on the performance of all operating pressurized water reactors.
Ilodels for treating the effects of fuel rod bowing on thermal-hydraulic performance hive been derived
~
These models are based on tlie propens'ity of the individual fuel designs to bow and on the
'hermal analysis methods used to predict the coolant condi tions for both normal operation and anticipated transients.
As a result of these evaluations the staff has concluded that in some cases sufficient thermal margin does not now exist.
In tlies>> cases, additional thermal margin will be required to assure, with hiali confidence, that departure from nucleate boiling (DNB) does not occur durinu anticipated transients.
This report discusses how these conclusions were reactied and identifies the amount of additional margin required.
The models and the required DNBR reductions which result from these models are meant to be only an interim measure until more data are available.
Because the data base is rather sparse,,
an attempt was made to treat this problem.in a conservative way.
The required DflBR reductions wi 11 be revised as more data become available.
g 4
~'he staff review of the amount and consequences of fuel rod bowing in a boiling water 'reactor i~ now underway.
At present no conclusions have been reached.
Hhen this review reaches a stage where either an interim or final conclusion can be reached, the results of this review will be published in a separate safety evaluation report.
It should be noted that throughout the remainder of this report, all discussion and conclusions apply only to pressurized water reactors.
C i ~
3 2.0 DNBR Reduction Due To Rod Bovi 2.1
~Back round In 1973 Westinghouse Electric presented to the staff the results of experiments in which a 4x4 bundle of electrically heated fuel rods was tested to determine the effect of fudl rod bowing to contact on the thermal margin(DNBR reduction)
(Reference 2),
The tests were done at conditions representative of PWR coolant conditions, The results of these experiments showed that, for the highest power density at the highest coolant pressure expected in a Westinghouse reactor, the DNBR reduction due to heated rods bovled to contact vias approximately 85.
Fuel bundle coolant mixing and heat transfer computer programs such as COBRA IIIC and THINC-IY viere able to accurately predict the tesul of these exper iments.
Because'he'end point could'be predicted, i.e., the DNBR reduction at contact, there was confidence that the DNBR reduction due to partial bovi, that is, bow to less than contact could also be correctly. predicted.
On August 9, 1976 Westinghouse met with the staff to discuss further experiments viith the same configuration of fuel bundle (4x4) using electrically heated rods.
However, for this set of experiments one of the center 4 fuel rods was replaced by an unheated tube of the same size as a Westinghouse thimble tube.
This new test configuration was tested over the same range of power, flow and pressure as the earlier tests.
However, with the unheated, larger diameter rod the reduction in DNBR was much larger than in the earlier (1973) tests.
Thg data consisted of poinLs corresponding to no intentional bowing (that is, a certain amount of'owing due to tolerances cannot be prevented) and to cont",ct.
No data were tat:
n at partial clearance reductions between rods.
The staff attempted to calculate the Westinghouse results with the COBRA IIIC comouter code but could not obtain aqreement with the new data.
Mestinghouse was also unable to obtain agreement I
between their experimental results and the THINCIV computer code.
On August 19, 1976 CE presented results of similar experiments to the staff.
These tests were performed using a
21 rod bundle of electrically heated rods and an unheated guide tube.
Results were presented for not only the case of,full contact, but.also the case of partial bowing.
Both sets of data (llestinghouse and CE) showered similar effects due to variations in coolant conditions.
For both cases, the DNBR reduction became greater as the coolant pressure and the rod power increased.
Because both sets of data showed that plant thermal margins might be less than those intended, the staff derived an interim model to conservatively predict the DNBR reduction.
Since the data with unheated rods could not be predicted by existing analytical
- methods, empirical models were derived.
These models give the reduction in DNBR as a function of the clearance reduction between adjacent fuel rods.
Two such models were derived, one based on the Westinghouse data and one based on the CE data.
~
~
t C)
- twW+ '<
2.2 l<odel Based on Mes~tin house Data As stated in Section 2.1, data were presented by Ilestinghouse for the ONBR reduction at full contact and with no how.
No data at
~ - >>'.t qn z'>>>>
r r ".,
~O~
0-partial gap closure were present<<:d.
- proposed, and the staff accepted, a straight line interpolation between these two points as shown in Figure 2.1.
This approach is conservative if the ONBR reduction does not increase more rapidly than the straight line reduction shown in I
Figure 2.1.
Although the data for DNBR reduction due to rod bowing in the presence of an unheated fuel rod cannot be predicted by existing analytical
- methods, one would nevertheless expect that the actual behavior would more nearly follow the curved line also shown in Figure 2.1.
According to this curved line, the ONBR would be reduced gradually for small amounts of bow.
As the fuel rods (or fuel rod and unheated rod) become close enough so that there is an inter-
- action, the DNBR would decrease more. rapidly.
No physical mechanism has been postulated which would lead to sudden large decreases in the DNBR for small or moderate gap closur es.
Thus, the straight line approximation is believed to be an overestimate of the expected behavior.
Experience with critical heat flux tests also supports the assumption of a small reduction in DNBR for small amounts of fuel rod bowing.
Experimental measurements of critical heat flux done on test assemblies always have some amount of rod bowing.
This may be due simply to fabrication tolerances or to electromagnetic-attraction forces set up between electrically resistance heated rods which simulate fuel rods.
2.3 It should be noted that this behavior (little or no reduction in DNBR for small amount of bowing) is shown by Combustion Engineering data which became available to the staff after the Mestinghou'e model was derived.
The Combustion Engineering data is discussed in Section 2.3 and the model derived from this data is shown in Figure 2.2.
All manufacturers of reactor cores, including Westinghouse, include a factor in their initial core design to account for the reduction in DNBR that may result from pitch reduction from fabrication tolerances and initial rod bow.
The amount of this pitch reduction factor varies with the fuel design and the analysis methods which are used.
For any particular core this factor is not varied as a function of burnup.
In developing the interim rod bow penalties described in this report, it became apparent that the penalty should be a function of burnup since the magnitude of rod bow is a function of burnup.
However, to maintain existing thermal margins early in core life when only a small amount of fuel rod bow is anticipated, the initial pitch reduction factor was included until such time as the rod bow DNBR reduction became greater.
This is represented as the straight horizontal line on Figure 2.1.
Combustion En ineerin Model Combustion Engineering performed experiments to determine the effect of rod bowing on DNBR which included some cases in which the effect of partial bowing as well as bowing to contact was determined.
- Again, a straight line interpolation is used.
However, the point of zero DNBR reduction is not at zero clearance reduction but rather,'t an intermediate value of'learance reduction.
This is shown schematical'ly
2.4 in Figure 2.2.
The horizontal straight line, representing the initial pitch reduction factor is included as explained previously in Section.2.2 I
thodels for Babcock and >lilcnx and Exxon On August 17, 1975 representatives of Babcock and Wilcox met with the staff to discuss this problem.
Babcock and Wilcox did not present any data on the effects of rod bowing on DNBR.
They had previously presented data to the staff on the amount of bowing to be expected in Babcock and lJilcox 15xl5 fuel assemblies.
Because Babcock and h!ilcox had no data on the effect of rod bow on DNBR, the staff applied the lJestinghouse model to calculate the effect of rod bowing on DNBR for Babcock and klilcox fuel.
This is acceptable since the conditions of operation are nearly the same in pressurized. water reactors from both vendors and the fuel bundle designs are similar.
The amount of fuel rod bowing as a function of burnup was calculated using the Babcock and Wilcox 15xl5 fuel bundle data.
I Representatives of the Exxon Nuclear Corporation discussed the effects of fuel rod bowing in the presence of an unheated rod on DHBR with the staff on August 19, 1976.
Exxon has not performed DNB tests with bowed rods and thus has no data pertinent to this problem.
The first cycle of Exxon fuel has just been removed from H.
B. Robinson and the results of measurements on the magnitude of rod bowing have not yet been presented to the staff.
The effects of fuel rod bowino for Exxon fuel were evaluated on a plant by plant basis as discussed in Section 4.0'
2.5 Apr.~ication of the Rod Bow/Di!BR Hodel Using these empirical models, the staff derived DNBR reductions to be applied to both operating reactors and plants in the Operating License.review stage.
The procedure in applying these empirical models is as follows:
~Ste 1:
predict the clearance reduction due to rod bow as a function of burnup.
An expression of the form
= a+bq~BU cC Co is used where
,= fractional clearance reduction due to rod bowing wC C
a,b
= empirical constants obtained for a given fuel design BU = burnup (region average or bundle average, depending on th' fuel designer).
Westinghouse showed in Reference 6 that an equation of the above form fit the rod bow data from 26 fuel regions'he constant a
represents the initial bow of the fuel rods due to fabrication tolerance.
The staff has approved the above equation (Reference 8).
Also included in the constants a and b is a factor of 1.2 to convert from the cold conditions at which the measurements were made to the hot operating conditions and a factor of 1.645 which, when multiplied by the standard devi.ation, gives an amount of bow greater than that expected from 95% of the fuel rods with a 95% confidence.
Step 2: Apply the previously discussed empirical models of BABB reduction as a function of clearance reduction using the value of lptC/co calculated from step l.
mes t>> n E k I
S~te 3:
'T!>e staff has permitted the reduction in DIIDP. calculated in step 2 to be offset by certain available therma] margins, These may be either generic to a given fuel design or plant dependent.
An example of a generic thermal margin which would be used to offset the DHBR reduction due to rod bow is thk fact that the DHBR limit of 1.30 is usually greater than the value of DHBR above which 955 of the data lie with a 95/ confidence.
The difference between 1.30 and this number may be used to offset the DHBR reduction, For 1lestinghouse 15xl5 fuel, the value of DHBR which is greater than 95K of the data at a 95% confidence level is 1.24 (Reference 1).
I For llestinghouse
'l7xl7 fuel this number is 1,28 (Reference 1).
A review of the data used to derive these numbers shows that the use of three signiificant figures is justified.
An example of a plant specific thermal margin would be core flow greater than the value given in the plant Technical Specifications.
A discussion'f the application. of this method to Construction Permit and Operating License reviews is given in Section 3.0.
A discussion of the application and the results of this method to operating reactors is given in Section 4.0.
The application to reactors using Exxon fuel is also discussed in Section 4.0.
hht" h r
E
~
t ~
t-3.0 3.1 plication to Plant in Construction Permit And ~0n~eratin License Review Stage
~CP A nl i ca ti on s No interim rod bow DNB penalties should be applied.to CP applications.
The rod bow data upon which the interim limits have been based should be considered preliminary.
There is sufficient time available to review the data and assess a penalty, if any, prior to the OL stage.
He will advise each CP applicant of the nature of interim penalties being applied to OL reviews and operating reactors.
As stated
- above, the data used to evaluate the effects of rod
-.rt tth+
t 4
3.2 bow on DNBR are preliminary.
They are also incomplete.
In order to III assess the conservatism of the straight line approximation and to, obtain data on designs for which no data is now available we will require the applicant to (1) fully define the gap closure rate for prototypical bundles and (2) determine by an appropriate experiment the DNB effect that bounds the gap closure from part (1).
Such requirements will be part of our CP review effort.
OL Ap lications Plants which are in the operating license review stage should eratcmtr ttsa consider a rod bow penalty.
This penalty should be as described in Section 2.2 for Hestinghouse or Section 2.3 for Combustion Engineering.
Babcock and Hilcox plants should use the rod bow vs.
burnup curve appropriate to their fuel and the Hestinghouse curve of DNBR reduction as a function of rod bow.
"i
~u ~c
~
~
I
~
gt
- ll-All applicants may propose appropriate thermal margins (as discussed in Section 2.4) to help offset the calculated DNBR reduction.
I
~ 'l
4.0 Ainlication To ~0erating Reactors This section divides the operating plants into distinct categories and lists them according to the fuel and/or reactor manufacturer, Operating plants which cannot be so categorized (such as plants with fuel supplied by more than one vendor) are placed in a separate category.
The plants assigned to each category are listed in the appropriate subsection.
The conclusions reached in this section are in some cases dependent on conditions or analysis which are valid only for the present fuel cycle.
- Hence, the FhH or DNBR reductions which are given (or the fact that no such reduction is concluded to bel required) is valid only for the present operating cycle.
4.1 Hestin house LOPAR Fuel The designation LOPAR stands for low parasitic and refers to the fact that the guide tubes in the fuel bundle are made of Zircaloy.
Table 4.1 gives a list of the operating plants which fall into this
=
classification.
TABLE 4.1:
PLANTS MHICH CURRENTLY USE THE WESTINGHOUSE LOPAR FUEL.
ASSEMBLY 15 x 15 17 x 17 Zion 1 Cycle 2
Zion 2 Cycle 1
Indian Point 3 Cycle 1
Turkey Po'int 3 Cycle 4
Turkey Point 4. Cycle 3 Prairie Island 2 Cycle 2
Prairz Island 1 Cycle 2
Trojan Cycle 1
Beaver Yalley 1 Cycle 1
TABLE 4.1 (cont.)
15 x'15 Surry 1 Cycle 4 Surry 2 Cycle 3 Kewaunee Cycle 2 Point Beach 1 Cycle 5
Point Beach 2 Cycle 3
I The reduction in Of(BR due to fuel rod bowing is assumed to vary, linearly with the reduction in clearance between the fuel rods (or fuel rod and thimble rod) according to the model discussed in Section 2.2.
The maximum value of DNBR 'reduction (at contact),
obtained from the experimental data was used to calculate the DNBR reduction
.vs.
bow for the 15xl5 LOPAR.fuel.
This DHBR contact reduction was.
adjusted for the lower heat flux in the 17x17 LOPAR fuel.
The clearance reduction is conservatively assumed to be given by the following equation for the 15xl5 (and 14xl4) fuel, a + bfBu aC Co is tr~e reduction in c1earance Co Bu is. the region average burnup and a,b are empirical consta'nts fitted to Westinghouse 15xl5 rod bow data
- 14 For the 17xl7 LOPAR fuel, the clearance reduction was calculated from the equation:
hC/Co
=
(aC) 15xlu 15xl5 17xl7 where L = the distance between grids I = moment of inertia of fuel rod s
m ls as smMmkisus
<ffusion coefficient) is a measure of the amount of mixinu between ad.iacent subcl(annels.
S~fCMa'>>
b$ga
~
s
~
TABLE 4.2:
FaH REDUCTION FOR MESTINGHOUSE LOPAR FUEL a
C'7 CYCLE REDUCTION IN FaH
{X)
.1 5x15 17x17 ZION 182
' a s+lj 1st Cycle
=(0-15 Gwd*/HTU) 2nd Cycle
{15-24 Gwd*/NTU) 3rd Cycle (24-33 Gwd*/NU) 0-2 ramp 0-9.5 12 12 0-6 ramp 10 These reductions in FIA-8 may be treated on a region:by region basis.
If the licensee
- chooses, credit may be taken for the margin I
between the actual reactor coolant flow rate and the flow rate used ir, I'..
r" 'dsi
(
4.2 safety calculations.
Credit may also be taken for a difference between the actual core coolant inlet temperature and that assumed in safety analyses.
In taking credit for coolant flow or inlet temperature
- margin, the associated uncertainties in these quantities must be taken into account.
llestinehouse HIPAR and Stainless Steel Clad Fuel The designation HIPAR stands for high parasitic and refers to the fact that the guide tubes in the fuel bundle are made of stainless steel.
These two fuel types, HIPAR and Stainless Steel clad, are grouped together because the amount of bowi ng expected
{and observed) is significantly less than that in the observed Westinghouse LOPAR fuel.
The plants which fall under this classification are listed in Table 4.3.
- Cud Hwd
= 1000 FTU t1TU.
TABLE 4.3:
HIPAR AND STAINLFSS STELL PLANTS Ginna Indian Point 2
San Onofre Connecticut Yankee The model for the reduction in DNBR due to fuel rod bowing is assumed to be identical to that used for the LOPAR fuel.
This is acceptable since cladding material should have no effect on CHF (critical heat flux) and the same DNB correlation applies to both HIPAR and LOPAR grids.
For reactors in this category, the peak reduction in DNBR (corresponding to 100% closure) was adjusted to correspond to the peak overpower heat flux of that particular reactor.
The amount of rod bowing for the plants listed in Table 4.3 which use HIPAR and stainless steel fuel, was calculated by means of an adjustment to the 15xl5 LOPAR formula.
This adjustment took the form of the ratio amount of bow for assembl~t
>e
= gL/IF) assy type
~L/I~E LOPAR L is the span length between grids amount of bow for LOPAR fuel where I is the moment of inertia of the fuel rod Ginna Cycle 6
E is the modulus of elasticity of the fuel rod cl adding The Ginna plant is fueled with
.121 fuel assemblies.
Two of these, are Exxon assemblies, and two are BSH assemblies.
The remainder are
'llestinghouse HIPAR fuel assemblies.
The experimental value of DNBR reduction was adjusted for heat flux and pressure from peak experimental to actual plant conditions.
Ginna took credit for the thermal margins due to pitch reduction, design vs. analysis values of TDC and
WW hP AF fuel densification pow:-.r spike.
These thermal margins offset the calculated DNBR reduction so that no reduction in FaH is required.
San Onofre Cycle 5
San Onofre is fueled with 157 bundles of 15xl5 stainless steel clad fuel, An FhH of 1.55 was used in thermal pesign and in the Technical Specifications.
To offset the reduction in FhH due to rod bowing San Onofre has proposed taking credit for margin available from the assumed worst case axial power distribution used in the thermal
'I analysis for San Onofre and that which would be possible during operation.
This proposal is now being reviewed by the staff.
Indian Point 2
Cycle 2
I Indian Point 2 is fueled with HIPAR fuel bundles.
The experimental value of DNBR reduction was adjusted for heat flux and pressure to actual plant conditions.
Indian Point Unit 2 had thermal margin to offset this DNBR reduction in pitch. reduction, design'vs.
analysis values of TDC, fuel densification power spike and a value of FaH of.
1.65 used in the design (vs. 1.55 in the Tech Spec).
Therefore, no, reduction of FsH is required for Indian Point Unit 2.
Connecticut Yankee Cycle 7
Connecticut Yankee is fueled with 157 stainless steel clad fuel assemblies.
e re u
Th DNBR d ction at contact was assumed to be that.
used-for the Westinghouse LOPAR 15'xl5 fuel.
No adjustment was made for heat flux.
e va ue Th 1
of pressure was adjusted to the overpress trip set point va ue o
1 f 2300 psi.
Full closure will not occur in '
stainless steel fuel,out to the design burnup.'onnecti cut an.ee as Y
k h
sufficient thermal margin i n variable overpressure an overp d
power trip set points to accommodate the calculated DNBR reduction.
Therefore no penalty is required.
FR, 4.3 Babcock and Wilcox 15xl5 The reactors listed in Table 4.4 are fueled with B&W fuel.
TABLE 4.4:
REACTOR USING B&W FUEL Oconee 1 Cycle 3
Oconee 2 Cycle 2
Oconee 3 Cycle 1
Rancho Seco Three Nile Island 1 Cycle 2
Arkansas 1 Cycle 1
Babcock and Wilcox met with the staff on September 8, 1975 and presented data on the amount of rod bow in B&W fuel.
The staff derived a model for B&M 15x15 fuel based on this data.
This model has the form:
aC
= a + b~BU Co where aC is the fractional amount of closure Co Bu is the bundle average burnup.
and a,b are empirical constants fitted to B&M data The reduction in DNBR due to fuel rod bowing is assumed to vary linearly with the reduction in clearance between the fuel rods {or fuel.
rod and thimble rod) but can never be lower than that due to the pitch reduction factor used in thermal analysis, as explained in Section 2.2.
Babcock and Wilcox claimed and the staff approved credit for the following,thermal margins:
. Flow Area (Pitch) reduction
. Available Vent Valve credit
. Densification Power Spike removal Excess Flow over that used in safety analyses
. Higher than licensed power used for plant safety analyses
Based on this review and.the thermal margins presented by B8W to offset the new Westinghouse
- data, Rancho Seco is 'the only plant for which a reduction in DNBR is required.
Table 5 gives the values for the reduction of DNBR required at this time.
TABLE 5:
DNBR REDUCTIONS FOR 88M PLANTS Burnup DNBR Reduction Pl Rancho Seco l
J Gwd Cycle 1 (0-15 ~j
)
Gwd Cycle 2 (15-24 gTU )
Gwd Cycle 3 (24-33
~~TU
)
0 1.6$
3Ã 4.4 Plans must be submitted to the staff to establish how these reductions in DNBR will be accommo'dated.
Combustion En ineerina 14x14 Combustion Engineering has presented data to the staff on the amount of'od bowing as a function of burnup. (Reference 5)
The staff used this data to derive the following model for CE 14xl4 fuel (Reference 7) a
+
b Bu, aC Co aC/Co
.-. fraction of closure for CE fuel Bu is the bundle average burnup
=M and a,b are empirical constants fitted to CE data
c
~c *cm 4 CE was given credit for thermal margin due to a multiplier of'.065 on the hot channel enthalpy rise used to account for pitch reduction due to manufacturing tolerances.
Table 4.6 presents the.
required reduction in DNBR using the model described above, after accounting for this thermal margin.
Table 4.7 is a list of the reactors to which it applies.
A licensee planning to operate at a burnup greater than 24000 Mwd/NTU should present to the staff an acceptable method of accommodating the thermal margin reduction sho>n in Table 4.6.
This may be done as part of the reload submittal if this burnup will not be obtained during the current cycle.
TABLE 4.6:
EFFECT OF ROD BOWING ON DNBR IN REACTORS WITH COMBUSTION ENGINEERING 14xl4 FUEL BURNUP Cycle 1 (0-15 ~ )
Gwd Cycle 2 (15-24 Gwd)
Cycle 3 (24-33 Gwd
)
TABLE 4.7:
PLANTS FUELED BY 4.6 APPLY St.
Luci e 1
Ft. Calhoun Hillstone 2
l1aine Yankee REDUCTION IN DNBR 0
0 CE FUEL TO WHICH VALUES OF TABLE Cycle 1
Cycle 3
Cycle 2
Cycle 2
Calvert Cliffs 1 Cycle 1
'.5 Plants Fueled Partiall With Exxon Fuel Palisades, H.
B. Robinson, Yankee Rowe and D.
C.
Cook are partially fueled with Exxon fuel.
A discussion of these reactors follows.'
Palisades Cycle 2
The Palisades reactor for Cycle 2 is fueled with 136 Exxon fuel.
assemblies and 68 Combustion Engineering fuel assemblies, The Combustion Engineering fuel was treated according to the Combustion Engineering model for both extent of rod bow as a function of burnup and DNBR reduction due to clearance reduction.
The Exxon fuel was assumed to bow to the same extent as the Combustion Engineering fuel, This assumption is acceptable since I
the Exxon fuel has a thicker cladding and other design features which should render the amount of bowi'ng no greater than in the Combustion Engineering fuel, The DNBR reducttiin was assumed to be linear with clearance reduction according to the Westinghouse type curve of Figure 2,1, The DNSR reduction at contact was based on the Westinghouse experimental data adjusted for the peak rod average heat flux in Palisades and for the coolant pressure jn Palisades, The variation of the DNBR reduction wi,th coolant pressure is given in Reference 1;
The DNBR reduction decreases as the coolant pressure decreases.
The overpressure trip set point in Palisades is set at, 1950 psi.
At this pressure, according to the data presented in Reference 1,
the penalty is greatly reduced compared to the penalty at high pressures.
The limiting anticipated transient in the Palisades reactor results in a DNBR of 1.36.
The thermal margin between this value and the DNBR limi.t of 1.3 results in adequate thermal margin to offset. the rod bow penalty.
'Yankee Rare CyÃe 12 I
Yankee Rowe is fueled with 40 Exxon fuel assemblies and 36 Qulf United Nuclear Corporation fuel assepbltes, The fuel assemblies consist of 16x16 2ircaloy clad fuel rods.
The reduction in DNBR due to fuel rod bowing was assumed to vary
'iinearly with the reduction in clearance between fuel rods, The peak experimental conditions used in the Westinohouse test were used to' fix the penalty at full closure, The calculated reduction in DNBR is still less than that which would produce a
DNBR less than 1,3 for the most limiting anticipated transient (two pump out of four pump loss-of-flow).
- Thus, no penalty is required.
H, B 'Robinson Cycle 5 H,
B, Robinson is fueled wi th 105 Westinghouse fuel assemblies and 52 Exxon Nuclear Corporation fuel assemblies, The Westinghouse
'5xl5 DNBR penalty model was applied to the westinghouse fuel with a correction for the actual heat flux rather than the peak experimental values.
The Exxon fuel was considered to bow to the same extent as the Westinghouse 15xl5 fuel so that the Westinghouse bow vs.
burnup equation was also applied to the Exxon fuel.
This assumption is conservative since the Exxon fuel has a thicker cladding and other desi'gn features which should render the amount of bowing no greater than in the Westinghouse fuel.
The DNBR reduction calculated by this method was offset by the fact that the worst anticipated transient for H.
B. Robinson results in a DNBR of 1.68.
C 3 D.
C.
Cook Cycle 2
D.
C.
Cook contains 128 Westinghouse fuel assemblies and 65 Exxon fuel assemblies.
The limiting transient for D.
C.
Cook is the Loss of Flow {4 pump coastdown) which has a minimum DNBR of 2,01, This value of DNBR is sufficiently high to accommodate the rod bow penalty for Cycle 2 without reducing the DNBR below the safety limit value of 1.3.
.-.24-5.0 References 1.
Letter to V. Stello, Director, Division of Operating Reactors, USNRC from C. Eicheldinger,
- Manager, Nuclear Safety Department, Hestinghouse Electric Corporation, NS-CE-N61, August 13, 1976.
2.
Hill, K. H. et., al, "Effects of a Bowed Rod on DNB", Hestinghouse Electric Corporation",
HCAP 8176.
3.
Standrad Review Plan - Section 4.4, II.l.A.
4.
Letter to R. Salvatori,
- Manager, Nuclear Safety Department, Hestinghouse Electric Corporation from D. Vassallo, Chief, Light Hater Reactors Project Branch l-l, Directorate of Licensing, December 4, 1974.
5.
Letter to V. Stello, Director, Division of Operating Reactors, USNRC, from P.
L. McGill, Combustion Engineering
- Company, December 15, 1975.
6.
- Reavis, J. R., et. al., "Fuel Rod Bowing" WCAP 8691 (Proprietary)
Hestinghouse Electric Corporation,
- December, 1975.
7.
Letter to Mr. Ed Sherer, Combustion Engineering from D.
F.
- Ross, Assi stant Director, Reactor Safety, May 14, 1976.
8.
Interim Safety Evaluation Report on Hestinghouse Fuel Rod Bowing Division of System Safety, USNRC, April, 1976.
25
++j+fi yr...
FIGURE 2.1 V/ESTfNGHOUSE lMODEL DNBR
~~~J'g r year'.(r
'*.'T~~TT "r r
K CQzOz z0 ODO LuK W CURVE EXPECTED BEHAVIOR THERMALDESIGN PENALTY INCLUDED IN ORIGINALDESIGN FiW 0
100%
hC Cp
FI GURE 2. 2 COMBUST)ON ENGIN".ERlNG MODEL DNBR CC COzDz z0I-ODD LLI lC CE CURVE~
THERViALDESIGN PENALTY INCLUDED IN INITIALDESIGN 0
1P0%
hC Cp