ML19345B421
| ML19345B421 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 11/21/1980 |
| From: | Howard J BOSTON EDISON CO. |
| To: | Ippolito T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19345B422 | List: |
| References | |
| 80-295, NUDOCS 8012010233 | |
| Download: ML19345B421 (14) | |
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% 9 BOSTON EDISON CCMPANY 800 SovLsToN 57 ACCT 80sicN. MAssAcMustTTs Q2199
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BEco. Ler. 480-295 ow; fis w+
ProposedChang#80-633 &rifa [j --
"Il Mr. Thomas A. Ippolito, Chief 6*
4 s jl Operating Reactors Branch #3 Division of Operating Reactors Office of Nuclest Reactor Regulation U.S. Nuclear Regualtory Commission Washington, D. C. 20555 License Fo. DPR-35 Docket No. 50-293 Proposed Technical Specification Change Concerning Single Loop Operation
Dear Sir:
Facility Operating License No. DPR-35 for Pilgrim Nuclear Pewer Station Unit #1 (PKPS-1) requires that the plant be shutdown if an idle recirculation loop can-not be returned to service within 24 hours. Boston Edison requests that this license provision be relaxed to allow operation at less than or equal to 50% power with one recirculation loop out of service and under the special operating conditions given in Section 2 below. The evaluation of this proposed mode of operation provided to Boston Edison by GE and described below, supports the conclusion that this mode of operation will not reduce safety margins. (1) Reason for Change The loss of a single recirculation pump has occurred at 'several operating BWR'a and is not, therefore, an improbable event. While the time required to procure necessary parts and to repair the loop depends on the nature of the failure, any loss of operating capacity would have a significant econ-omic effect. Modifying the Technical Specifications as requested will reduce this potential economic impact without reducing the safety of plant operation. (2) Special Operating Conditions for Single Loop Operation In order to ensure that operation in this derated condition is in accordance with ! the assumptions utilized by GE, BECo commits to the following conditions during normal operation. Technical Specifications incorporating these con-ditions are included as Attachment A to this letter. R012cleN3 9
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- EOCON COMPANY Mr. Thomas A. Ippolito, Chief November 21, 1980 fage 2 A. The idle recirculation loop recirculation pump is electrically disarmed
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and the motor is inoperable precluding operation of the pump or injection
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of a cold slug into the vessel.
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B. The recirculation controls will be placed in the manual mode, thereby eliminating the need for control system analyses. C. The settings for the rod block monitor, APRM rod block trip, and flow bias scram will be modified as necessary to provide for single loop operation. D. MAPLHGR restrictions will result in a 35 percent reduction for all fuel. E. BECo vill limit the power level to 50%.
- 3. MAPLHGR Adjustment Factor for PNPS General Electric is performing analyses for single loop operation of PNPS-1.
Preliminary evaluation of these calculations performed according to the procedure outlined in NEDO-20566-2 indicatec that a multiplier of 0.83 should be applied to the MAPLHGR limits for single loop operation of the PNPS-1. Further, GE has performed a large number of single loop analyses for similar plants; in no case has a multiplier of less than 0.70 been required. Because PNPS-1 does not have LPCI modification and because the limiting break is a suction line break, the single loop MAPLHGR multiplier is expected to be significantly better than for most other BWR's. Until the PNPS-1 calculations can be verified (as required by 10CFR50 Appendix B), it is proposed that a multiplier of 0.65 be conservatively applied for single loop operation at the PNPS-1.
- 4. Safety Considerations for Single-Loop Operation Various conditions have been examined for the impact of single-loop operations.
The following pages (Attachment A) address several issues including: A. One-pump Seizure Accident B. Abnormal Operational Transients
- 1. Transients and Core Dynamics
- 2. Rod Withdrawal Error
- 3. APRM Trip Setting
- 4. Kg Curves
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C. Stability Analysis l D. Thermal - Hydraulics
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' COOTQN EOCON COMPANY Mr. Thomas A. Ippolito. Chief November 21, 1980 Page 3
- 5. Schedule of Chance
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t This change will be put into effect upon receipt of approval from the
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- 6. Tee Consideration In accordance with Section 170.12 of the Commission's Regulations, Boston Edison proposes this change as Class III. Accordingly a check for four thousand ($4,000) is, enclosed.
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_ Should you have any questions on this subject, please do not hesitate to contact us. Very tru? / yours ,
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gy (@ Attachments U 3 signed originals and 37 copies Commonwealth of Massachusetts) County of Suffolk ) Then personally appeared before me J. Edward Howard, who, being duly sworn, did state that he is Vice President - Nuclear of Boston Edison Company, the applicant herein, and that he is duly authorized to execute and file the sub-mittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.
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My Commission expires: fu!y by /90U l]b4baj h! e)) Nothry Pub 1Tc' /
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ATTACHMENT A (SAFETY CONSIDERATIONS) ONE-PLHP SEIZURE ACCIDENT The pump seizure event is a very mild accident in relation to other accidents such as the LOCA. This has been demonstrated by analyses in (NEDE-24011)* for the case of two-pu=p operation, and that it is also true for the case of one-pbmp operation is easily verified by consideration of the two events - In both accidents, the recirculation driving loop flow is lost extremely ranidly: in the case of the seizure, stoppage of the pump occurs; for the LOCA, the sever-ance of the line has a similar, but more rapid and severe influence. Following a pump seizure event, natural circulation flow continues, water level is maintained, the core remains submerged, and this provides a continuous core cooling mechanism. however, for the LOCA, complete flow stoppage occurs and the water level decreases due to loss-of-coolant resulting in uncovery of the reactor core and subsequent
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overheating of the fuel rod cladding. .In addition, for the pump seizure accident, < reactor ' pressure does not decrease, whereas complete depressurization occurs. for ' dhe LOCA. Clearly, the increased temperature of the cladding and reduced reactor pressure for the LOCA both combine to yield a much more severe stress and potential ' for cladding perforation for the LOCA than for the pump seizure. Theref ore, it can be concluded that the potential effects of the hypothetical pump seizure accident are very conservatively bounded by the effects of a LOCA and specific analyses of the pump seizure are not required.
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* " Generic Reload Fuel Application" NEDE-24011-P-A July 1979
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. . ABNORMAL OPERATIONAL TRANSIENTS TRANSIENTS AND CORE DYNAMICS Since operation with one recirculation loop results in a maximum power output which ir 20 to 30% below that from which can be attained for two-pump operation,
, ,the consequences of abnormal operational transients from one-loop operation will be considerably less severe than those analyzed from a two-loop operational mode.
For pressurization, flow decrease, and cold water increase, transients previously transmitted for Reload /FSAR results bound both the thermal and overpressure con-sequences of one-loop operation. Figure I shows the consequences of a typical pressurization transient (turbine trip) as a function of power level. As can be
.seen,, the consequences of one-loop operation are considerably less because of the associated reduction in operating power.-level.
The consequences from flow decrease transients are also bounded by the full power analysis. A sir- e pump trip from one-loop operation is obviously less severe than a two-pump t;." from full power because of the reduced initial power level. Cold water increase transients can result from either recirculation pump speed-up or introduction of coldar water into the reactor versel by events such as loss of feedwater heater. For the former, the Kf facters are derived assuming that both recirculation loops increase speed to the maximum permitted by the M-G Set scoop tube position set screws. This condition produces the maximum possible power increase and hence maximum AMCPR for transients initiated from less than rated power and flow. When operating with only one recirulation loop, the flow and power increase associated with the increased speed on only'one M-G Set will be less than that associated with both pumps increasing speed, and therefore, the Kg factors derived with the two pump assumption are conservative for single-loop operation. For the latter, the loss of feedwater heater event is generally the most severe r.old water increase event with respect to increase in core power. This event is caused by positive reactivity insertion from core flow inlet subcooling; therefore, the event is independent of two-pump or one-pump operation. The severity of the event is primarily dependent on the initial power level. Thu higher the initial power level, the greater the CPR change during the transient. Since the initial power level during one-pump operation will be significantly lower, the one-pump cold water increase case is conservatively bounded by the full power (two-pump) analysis. From the above discussions, it can be concluded that the transient consequence from one-loop operation is bounded by previously submitted full power analysis. The maximum power level than can be attained on one-loop operation is only re-stricted by the MCPR and overpressure limits established from a full power analysis. ROD WITHDRAWAL ERROR The rod withdrawal error at rated power is given in reload licensing submittals. These analyses demonstrate that even if the operator ignores all indications and alarm which could occur during the course of the transient, the rod block system will stop rod withdrawal at a critical power ratio which is higher than the safety 1 Lait MCPR. The MCPR requirement for one-pump operation will be equal to that for two-pump operation because the nuclear characteristics are independent of whether the core flow is attained by one- or two-pump operation. The only exceptions to th,is independence are possible flow asymmetries which might Page 2 of 9
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. . result from one-pump operation. Flow asymmetries are shown to be of no concern _ by tests conducted at Quad Cities. Under conditions of one-pu=p operation and
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equalizer valve closed, flow was found to be uniform in each bundle. One-pump operation results in backflow through 10 of the 20 jet pumps while the flow is being supplied into the lower plenum from the 10 active jet pumps. Eaca'use "of the backflow through the inactive j et pumps, the present rod block
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equation shown in the Technical Specification must be modified.
- The procedure for modifying the rod block equation for one-pump operation is given in the following subsecticns.
- a. The two-pump rod block equation in the existing Technical Specification is of the form:
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RB = (mW + K)% where RB = power at rod block in % m = flow reference slope for the rod block monitor (RBM) W = drive flow in % of rated K = power at rod block in % when W = 0. For the case of top level rod block at 100% flow, denoted RB100' RB 100
= m(100) + K or K = RB 100 - m(100)
Substitutint for K in Equation (1), the two pump equation becomes: RB=mW+(RB 100 - m(100 (2)
- b. Next, the core flow (F ) versus drive flow (W) curves are deturmined for the two-pump and one-pump cases. For the two-pump case the core flow and drive flow are derived by measuring the differential pressures in the jet pumps and recirculation loop, respectively. Core flow for one pump operation must be corrected for the backflow through the inactive jet pumps thus:
Actual core flow (one pump) = Active jet pump flow - inactive jet pump flow. Both:the active and inactive flows are derived from the jet pump differential pressures. The' drive flow is derived from the differ-ential pressure measurement in the active recirculation loop. These two curves are plotted for a typical BWR in Figur:i 1. The maximum difference between the ona-pump and two-pump c.fe flow is determined graphically. This occurs at about 35% drive flow which is denoted W. Page 3 of 9 Jg - - - .
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- c. Next, a horizontal 'line is drawn from the 35% drive flow point on the one is pump curve to the two pump curve and the corresponding flow, Wy ,
determined. Thus , AU = W y -W' 2 The rod block equation corrected for one pump flow is:
- - RB = mW + RB - m(100) - A RB
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where ( A RB ~ = RB - RB 2 *aw 7 RB = mW + RB 100 - m(100 + AW) (3)
- d. For_PNPS application', the constants from the Technical Specification are:
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.m = 0.65 RB = 107 100 i
From Figure 2: AW = W -W 2 = 35 - 30 = 5 3 Evaluating in Equation (3),the one-pump rod block equation becomes: (4) RB = 0.65W + 107 - 0.65(100+5) = 0.65W + 38.7 This line is depicted in Figure 2 as the future corrected rod block line for one-pump operation. APRM TRIP SETTING The APRM trip settings are flow biased in the same manner as the rod block monitor trip setting. Therefore, the APRM rod block and scram' trip settings are subject to the same procedural changes as the rod block monitor _ trip setting discussed above. i KgCURVE For single recirculation loop operation, the Kf curve contains sufficient con-servatism to provide operational limits such that the fuel integrity safety limit is not violated for abnormal operational events.
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- STABILITY ANALYSIS The least stable power / flow condition attainable under normal conditions occurs at natural circulation with the control rods set for rated power and flow. This
- condition may be reached following the trip of both recirculation pumps.
Operation along the minimum forced recirculation line with one pump running at minimum ' speed is more stabic than operating with natural circulation flow only, but is less stable than operating with both pumps operating at minimum speed. The. core stability along the forced circulation, rated rod pattern line for single loop operat an is the. same as that for both loops operable except that rated power ' is not attainable. Hence, the core is limited to maximum power for single pump operation and only manual flow control should be used. This is illustrated generically in Figure 3. .
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. . I THERMAL-HYDRAULICS Most.of the uncertainties used in the-statistical analysis presented in Table 5-1 of NEDE-24011 are independent of whether flow is provided by two-loop or s, ingle-loop. The only exception is the core total flow. The standard devia-tion for this quantity from Table 5-1 is 2.5%. For single-loop operation this value may increase to about 6%.of rated . core flow. The 3.5% increase in core total flow uncertainty corresponds to an increase in the safety limit of about
'O.01 which is more than of fset by the Kg factor required at low flows.
The steady-state operating MCPR with single-loop operation vill be conservatively established by multiplying the Kf factor to the rated flow MCPR limit. This ensures that the. 99.9% statistical limit requirement is always satisfied. ,
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