ML19261D921
| ML19261D921 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/29/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19261D915 | List: |
| References | |
| NUDOCS 7906290019 | |
| Download: ML19261D921 (11) | |
Text
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UNITED STATES
[
g NUCLEAR REGULATORY COMMISSICN E
WASHINGTON. D. C. 20005
\\...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENOMENT NO. 61 TO PROVISIONAL OPERATING LICENSE NO. DPR-21 NORTHEASTNUCLEARENERGYCOMPANY(ETAL)
MILLSTONE NUCLEAR POWER STATION', UNIT 1 DOCKET NO. 50-245 1.0 Introduction By letter dated March 5,1979, Northeast Nuclear Energy Company (NNECO)(thelicensee)requestedanAmendmenttoProvisional Operating License No. DPR-21 to change the Technical Specifications for the Millstone Nuclear Power Station, Unit No.1 (Millstone 1).
The proposed changes would allow use of 148.8x8R type fuel assemblies,
~
constituting refueling of the core for Cycle 7 operation (Reload 6) at power levels up to 2011 MWt (100% power).(Reference 1).
In support of the reload application, the licensee provided the GE BWR Reload Licensing Submittal for Reload 6, (Reference 2) proposed Technical Specification changes (Reference 3) and information on a regmented rod test bundle (Reference 4). Additional information on coastdown at end-of-cycle has also been provided (Reference 5).
This reload involves loading)of GE 8x8 and 8x8R fuel (plus some.
7x7 fuel from the first core. The description of the nuclear and mechanical design of the 8x8 and Ex8R fuel is contained in GE's licensing topical report for reloads (Reference 5) and a recent letter report (Reference 7). Reference 6 also contains a complete set of topical reports which describe GE's analytical methods for nuclear, thermal-hydraulic, transient and accident calculations, and infomation regarding the applicability of these methods to cores containing 7x7, 8x8, and 8x8R fuel.
Because of our review of a large number of generic considerations related to use of 8x8R fuel in mixed loadings with 8x8 and 7x7 fuel, and on the basis of the evaluation presented in Reference 6, only a limited number of additional areas of review have been included in this safety evaluation report. Evaluations not specifically addressed in this safety evaluation report, are identified in Reference 6.
2306 44 7 906290 d/7 2.0 Evaluation 2.1 Nuclear Characteristics During Cycle 7 operation of Millstone 1,100 fresh 8x8R bundles of type 80RB265L and 48 fresh 8x8R bundles of type 8DRB265H will be loaded into the core (Reference 2). The remainder of the 580 fuel bundles will be 8x8 fuel bundles exposed during previous cycles, plus 40 previously exposed 7x7 bundles in peripheral locations.
Based upon the data provided in Reference 2, both the control rod system and the standby liquid control system will have acceptable shutdown capability during Cycle 7.
2.2 Thermal-Hydraulics 2.2.1.
Fuel Cladding Integrity Safety Limit As stated in Reference 6, the Minimum Critical Power Ratio (MCPR) which may be allowed to result from core-wide or localized transients is 1.07.
This limit has been imposed to assure that during transients 99.9% of the fuel rods will avoid transition boiling.
The safety limit MCPR for Millstone 1 is being raised from 1.06 to 1.07 because the distribution of fuel rod power within the 8x8R fuel bundles is different from that of the 8x8 fuel. The reason for the difference is the presence of two rather than one water rod in 8x8R fuel. The issue has been addressed in Reference 6 and the 1.07 limit has been found acceptable for BWRs with uncertainties in flux monitoring and operational parameters no greater than those listed in Table 5-1 of Reference 6, for which the CPR distribution is within the bounds of Figures 5.2 and 5.2a of Reference 6.
It has been shown in Section 5 of Reference 6 that these conditions are met for Millstone 1.
2.2.2 Operating Limit MCPR Various transients could reduce the MCPR below the intended safety limit MCPR during Cycle 7 operation. The most limiting of these operational transients have been analyzed by the licensee to determine which event could potentially induce the largest reduction
' in the initial critical power ratio (ACPR).
The transients evaluated were the limiting pressere and power increase transient (in this case, the load rejectica transient without turbine bypass to the n.ain condenser), the limiting coolant temperature decrease transient (loss of feedwater heater),
the feedwater controller failure transient, and the control rod withdrawal error transient.
Initial conditions and transient input parameters used in the analysis are specified in Sections 6 and 7 of Reference 2.
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- The calculated systems responses and ACPRs for the above listed operational transients and conditions have been analyzed by the licensee.
Results were as follows:
TRANSIENT CALCULATED ACPR BY FUEL TYPE 7x7 8x8 & 8x8R Load Rejection without Bypass
.20
.27 Loss of a Feed-water Heater
.13
.15 Feedwater Controller Failure
.06
.07 Rod Withdrawal Error NA*
.15 Addition of the most severe ACPR to the safety limit (1.07) gives the appropriate operating limit for each fuel type. This will assure that the Safety Limit MCPR is not violated due to transients.
Using the above table, the licensee has proposed the following operating limits:
1.34 for 8x8 and 8x8R fuel 1.27 for 7x7 fuel Because these limits will preclude violations of the Safety Limit MCPR in the event of any anticipated operational occurrence, we find these limits to be acceptable.
.2.3 Accident Analysis 2.3.1 ECCS Appendix X Analysis Input data and results for the ECCS analysis have been given in References 5, 6, and 8.
The information presented fulfills the requirements for such analyses outlined in Reference 6.
We have reviewed the information submitted for the reload and conclude that Millstone Unit 1 will be in conformance with all
- requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50.46 when (1) it is operated within the " MAXIMUM AVEPAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VS. PLANAR AVERAGE EXPOSURE" values
- All 7x7 bundles are located on the core periphery. These locations are never analysed for rod withdrawal errors because of their lower power.
For this reason peripheral control rods are not connected to the Rod Block Monitor.
2306
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given in Figure 3.ll.lh of Reference 3 and similar figures for the exposed fuel already in the Technical Specifications, and (2) it is operated at a Minimum Critical Power Ratio greater than or equal to 1.20.
(More restrictive MCPR limits are currently required for reasons not connected with the Loss of Coolant Accident, as described in Section 2.2.2).
2.3.2 Control Rod Drop Accident The Rod Worth Minimizer (RWM) and associated rod pattern procedures at Millstone 1 use GE's Banked Position Withdrawal Sequence (BPWS).
Generic analyses for BWR/2 and BWR/3 plants using BPWS, described and sunnarized in Reference 6, have shown that the peak fuel enthalpy deposited during a rod drop accident will be less than the bounding analysis limit of 280 cal /gm provided the maximum incremental control rod worth is not greater than 1.0% AK.
The licensee has performed an incremental control rod worth compliance calculation for Millstone 1 reload 6, and found a maximum incremental worth of 0.78% AK (Reference 2).
This is well within the bounding analysis limit of 1.0% AK, and therefore we find the analysis to be acceptable.
2.3.3 Fuel Loading Error The licensee has examined the reloaded core and determined that the worst-case fuel loading error, mis-locating a new 8DRB265 bundle into a location intended for an 8DRB262 bundle, results in a MCPR of 1.09.
Since this is greater than the safety limit of 1.07, we find this analysis to be acceptable.
2.3.4 Overpressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been performed in accordance with the requirements of Reference 6.
As specified in Appendix C of Reference 6, the sensitivity cf peak vessel pressure to failure of one safety valve has also been evaluated. We agree that there is sufficient margin between the peak calculated vessel pressure and the design limit pressure to allow for the failure of at least one valve. Therefore, the limiting over-
. pressure event as analyzed by the licensee is considered acceptable.
2.4 Thermal-flydraulic Stability The results of the thermal-hydraulic stability analysis (Reference
- 2) show that the channel hydrodynamic and reactor core decay ratios at the natural circulation - 105% rod line intersection (which is the least stable physically attainable point of operation) are below the stability limit.
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. Because operation in the natural circulation mode will be prohibited by the Technical Specifications, there will be added margin to the stability limit and this is, therefore, acceptable.
2.5 Physics Startup Testf,rg The licensee has nce, changed his startup test program from that approved for the pr evious cycle. This program, therefore, remains acceptable.
2.6 Segmented Test Rod Bundle The licensee will load one special test bundle (Reference 4). This test bundle is used to investigate cladding behavior and has been irradiated in the previous cycle. Other than its special cladding, it is similar to, standard fuel in its mechanical and nuclear design, and is included with the other bundles in the various safety analyses evaluated herein. Therefore, operation with this test bundle in the core continues to be acceptable.
2.7 End-of-Cycle Power Coastdown The licensee desires to coast down to 70% of rated power after the end of the normal cycle (References 1 and 5).
Reactor thennal power will be extended by reducing feedwater temperature by 75 F.
As core exposure exceeds that corresponding to the all-rods-out full-power condition, the decreasing fissile inventory causes core reactivity to decrease. Under nonnal coastdown circumstances, power automatically decreases such that the reduced power reactivity defect exactly cc.mpensates for the effect of reduced fissile inventery.
In addition to this integral effect, ietailed core parameters are affected by power coastdown as foliows:
(a) The void fraction decreases.
(b) The coid coefficient generally becomes less negative.
(c) Peaking factors may change in either direction.
(d) Axial power distribution may shift in either direction initially, but ultimately will shift upwards.
degrades in (e) The scram reactivity insertion curve eventually (nearly fully its initial portion and improves in its final inserted) portion.
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=..
e The overpressure analysis and the transient analyses are affected by these parameters.
In cases where these parameters exceed the bounding values used as inputs to these analyses, re-analysis may be necessary.
In general, (a) and (b) improve transient response but (d) and (e) have the opposite effect.
The reduction of feedwater heating extends core life by inserting positive reactivity via the moderator coefficients.
It also has the effect of moving the axial power distribution lower in the core, thus countering effects (d) and (e).
The licensee has re-calculated the transients and the overpressuri-zation event at 70% power, assuming coastdown with 75*F feedwater reduction from an E0C Haling axial power' distribution (Reference 5).
Results are as follows:
TRANSIENT ACPR VESSEL PRESSURE 7x7 8x8 a 8x8R EOC Coastdown E0C Coastdown EOC Coastdown Load Rejection
.20
.16
.27
.21 1225 1206 Loss of 100 F Feedwater Water Heating
.13
.10
.15
.12 1072 1068 Feedwater-Contmiler Fai' -e
.06
.07
.07
.09 1072 1070 MSIV Closure (Overpressure) 1275 1245 It can be seen that the calculated peak vessel pressure decreases in each case, thus preserving both overpressure protection and simer margin. Moreover, the limiting transient remains the load rejection without bypass with a smaller ACPR for the end of core life reduced temperature power coastdown conditions.
The transient analyses described above were performed with the REDY code (Reference 9).
A new improved code, ODYN, has been developed by GE.
The ODYN code, which uses a more accurate model of the plant,
~ generally predicts smaller ACPRs than the REDY Code when the transient under study is fairly severe. However, as transient severity is lessened, ODYN predicts a greater ACPR than REDY (Reference 10, p.1).
Both codes are run with conservative
- The full power EOC values are presented for comparison.
2306 349
~
input values, but ODYN should be a better predictor of plant behavior once these inpJt values are specified.
GE has stated (Reference 10) that REDY can be used here because the limiting transient has a ACPR sufficiently large to be above the regior, where REDY is non-conservative with respect to ODYN. Reloads have been approved on this basis.
The reduced aCPR calculated under coastdown conditions may be subject to some nonconservatism since the ACPRs have become smaller.
Based upon the information in Reference 10, we estimate this nonconservatism to be at most.02 in ACPR for 8x8 and 8x8R fuel, and.04 for 7x7 fuel, for the limiting (load rejection) transient.
(Considerations of steam line dynamics at reduced power might reduce or eliminate these values, but such considerations are not necessary here).
Since the margin between the calculated coastdown ACPRs and the E0C ACPRs upon which the limits are based covers this possible nonconservatism, and in view of the limitations of the REDY code, we conclude that the proposed technical specifi-cation limits are acceptable.
Based on the above, we conclude that operation in the proposed coastdown mode is acceptable.
By " proposed coastdown mode,"
we mean operation with the feedwater temperature remaining reduced. We do not consider coastdown with reduced feedwater heating, followed by a return to operation with full feedwater heating, to necessarily be bounded by the licensee's analysis.
2.8 Technical Specifications The licensee has submitted proposed changes to the Millstone 1 Technical Specifications (Reference 3). The proposed changes would allow the licensee to (1) change the Safety Limit MCPR from 1.06 to 1.07 in accordance with Section 2.2.1 above, (2) change the Cycle 7 operating limit MCPRs in accordance with Section 2.2.2, (3) change the MAPLHGR limits in accordance with Section 2.3.1, (4) delete the dropped rod worth in Specification 3.3.B.3, and (5) revise the setpoint, maintenance and surveillance require-ments on the Safety / Relief Valves.
Of these, (1), (2) and (3) have already been discussed in previous sections and found acce table. The licensee has provided seperate justifications
. for 4) and (5) in Reference 3.
We shall discuss these in turn.
2306 350 2.8.1 Dropped Control Rod Reactivity Worth The licensee has proposed deleting a limit on dropped rod reactivity worth. We find this to be acceptable because the intent of this specification is already contained in the BPWS analysis (see Section 2.3.2) and Technical Specifications enforcing use of the Rod Worth Minimizer.
Moreover, deletion of this requirement is consistent with current practice, including the Standard Technical Specifications for GE Boiling Water Reactors (Reference 11).
2.8.2 Safety / Relief Valves The overpressurization analysis evaluated in Section 2.3.3, which was calculated using the revised Safety / Relief Valve (S/RV) set-points, concluded that overpressurization protection was adequate.
Moreover, the higher setpoints will increase the margin between nonnal operating pressure and S/RV setpoint pressures (sinner margin), thus reducing unnecessary opening of the valves and increasing their reliability. The licensee has also examined the effects of the increases S/RV setpoints on the torus and on the main steam and relief valve piping (Reference 3). This examination demonstrated that the forces on these ' components during S/RV actuations are bounded by analyses previously accepted by the NRC staff (Reference 12). Therefore, we find the revised setpoint specifications to be acceptable.
In addition, the licensee will replace the S/RVs with a new two-stage design (Reference 3). This change necessitates some changes to the S/RV surveillance requirements (e.g., deletion of bellows monitoring requirements, since the new valves have no bellows).
The revised surveillance requirements are consistent with those previously approved for other installations using the new two-stage valves (Reference 13). Therefore, we find them to be acceptabl e.
3.0 Environmental Considerations 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 involve an action which'is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 9 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.
2306 J51 4.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Date:
May 29, 1979 2306 352 i
e g
e
4.0 References 1.
Letter, W. G. Counsil (NU) to Director of Nuclear Reactor Regulation (NRC), dated March 5,197 9.
2.
" Supplemental Reload Licensing Submittal for Millstone Unit 1 Reload 6," NED0-24168, December 1979, enclosure of Reference 1.
3.
" Description of Technical Specification Changes" and " Proposed Technical Specification Changes," enclosures to Reference 1.
4.
"STR Bundle Submittal, Millstone Unit 1 Segmented Test Bundle,"
NEDE-20592-4P, Supplement 4, January 1979, Enclosure of Letter, W. G. Counsil (NU) to Director of Nuclear Reactor Regulation (NRC), dated March 5,1979 and Revision 1 April 1979 Enclosure of NNEC0 letter to NRC, dated May 2,1979.
5.
" Supplemental Reload Licensing Submittal for Millstone Unit 1.
Reload 6," NEDO-24168-1, Supplement 1. February,1979, enclosure to Reference 1.
6.
" General Electric Boiling Water Reactor Generic Reload Fuel Appli cation," NEDE-240ll-P-A, May,1977.
7.
Letter, R. E. Engel (GE) to Division of Operating Reactors, NRC, dated January 30, 1979.
8.
" Loss-of-Coolant Accident Analysis Report for Millstone Nuclear Power Station Unit 1," NE00-24085, December 197,7.
9.
" Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor," NED0-10802, February.
1973.
10.
" Impact of One-Dir.ensional Transient Model on Plant Operations Limits," enclosure of letter, E. D. Fuller (GE) to U.S. Nuclear Regulatory Commission, dated June 26, 1978.
11.
" Standard Technical Specifications for General Electric Boiling Water Reactors," NUREG-0123, August,1976.
a
- 12. Letter, B. K. Grimes and R.
H.. Vollmer (NRC) to All BWR Licensees (except Humboldt Bay, Big Rock Point, and Lacrosse), dated T
February 27, 1979.
13.
"Edwin I. Ha'-h Nuclear Plant Unit 2 Technical Specification - -
Appendix A to License No. NPF-5," NUREG-0395, dated June 13, 1977.
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