ML20062C251
| ML20062C251 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/24/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20062C243 | List: |
| References | |
| NUDOCS 7811070311 | |
| Download: ML20062C251 (5) | |
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'o, UNITED STATES f ',,,
j NUCLEAR REGULATORY COMMisslON 3j ~. y VIASHINGTON. D. C. 20555
\\',%.....j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 29 TO FACILITY OPERATING LICENSE N0. DPR-61 CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLANT DOCKET NO. 50-213 Introduction By letters dated September 22,1978 (Reference 1) and October 20, 1978, (Reference 2) Connecticut Yankee Atomic Power Company (CYAPCO) submitted an analysis justifying the assumption that, with offsite power unavailable, removal of charging pumps from the automatic sequencing initiated by the safety injection signal will not cause the reactor to return to criticality after its initial scram following a main steam line break accident (MSLB).
The licensee has proposed the following changes to the Technical Specifications (Reference 3):
(1) Imposing of 3% shutdown margin during subcritical operation, and (2) Reduction of the trip setpoint from 109% to 25% of rated power when the reactor power is less than 10% of rated power level.
We discussed with thelicensee a modification to the bases for the g
control rod insertion limits that specifically states that these limits provide sufficient shutdown margin to ensure that the reactor does not return to criticality for any postulated accident in Chapter 10 of the Facility Cescription and Safety Analysis (FDSA). The iicensee agreed to this modification.
In its previous submittal (Reference 4) the licensee requested to remove the charging pumps from automatic actuation by the safety injection signal whenever offsite power is not available. This change was necessary because the emergency diesel generator set was undersized and could not provide sufficient load capacity to operate all the safety injection equipment, including the charging pumps. The licensee has acknowledged that removal of the charging pumps could have some effects on the perfonnance of the safety injection system during two postulated accidents:
loss of coolant (LOCA) and main steam line break (MSLB).
In subsequent investigations it was revealed that the 7g1107 030
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, large and small break LOCA analyses presented in the FDSA (Reference 5) were performed without taking credit for the liquid injected by the charging pumps. These analyses are, therefore, applicable to the present case. However, the original FDSA analyses of steam line break did take l
l credit for the boron injected by the charging pumps. These analyses were also performed assuming offsite power available.
It was therefore necessary to perfom a new analysis to assure that the reactor would not return to criticality following its initial scram. The licensee had i
provided such an interim analysis which indicated that charging pumps 7,
are not required to mitigate the consequences of a main steam line break accident until at least 300 Effective Full Power Days (EFPD) into the present cycle (Cycle 8). Beyond this time the moderator temperature coefficient could become sufficiently negative (due to low boron
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concentration) to cause addition of positive reactivity during cooldown which eventually could exceed the shutdown margin.
1 By letter dated May 24,1978 (Reference 4) the licensee committed to provide, prior to 300 EFPD in Cycle 8, either a MSLB analysis supporting operation beyond 300 EFPD without automatic operation of the charging pumps, or to propose a design change which provide automatic operation of the charging pumps on the diesel generator set. On May 25, 1978 (Reference 6), we advised the licensee that these comitments should be incorporated into the facility operating license. The licensee concurred with the license condition.
Evaluation The licensee has perfomed the MSLB analyses assuming offsite power is not available and taking no credit for the boron injected by the charging
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pumps (References 1 and 2). The analyses were performed using the NRC approved RELAP 4/M00 5 computer code and assuming end of cycle conditions (EOC).
These conditions were chosen because they yielded the most negative moderator coefficient causing the most pronounced insertion of positive reactivity during moderator cooldown. The cases at hot zero power i
(Reference 1) and hot full power (Reference 2) were analyzed.
In addition the licensee has compared the existing with the required shutdown margins for several intemediate power levels (Reference 2).
In all of these cases the control rod configurations corresponding to the maximum allowable control rod insertion limit was postulated (Figure 3.10-1 in Technical Specifications). This assumption was conservative. We have also reviewed and found acceptable the metFodology accounting for the boron injection and the non-equilibrium condition in the pressurizer.
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The highest degree of moderator cooldown caused by a MSLB accident was found to occur at zero power due to the increased mass inventory in the steam generators and decreased stored energy in the primary system.
However, this effect was partially offset by the reduced reactivity and Doppler coefficients at low temperatures. The net effect was that the addition of positive reactivity was monotonically increasing with reactor power. The available shutdown margins due to the control rods also changed with reactor power. The primary reason for thiss change was the variation in the maximum allowable inserted control rod worth which was higher at lower reactor power and kept decreasing as power increased to full power and rods were withdrawn from the core. The value of the maximum stuck control rod also varied with power and was generally lower at lower power levels. The licensee has determined that the curve representing the existing shutdown margin exhibited a minimum at about i
40% of rated power level. However, it was clearly demonstrated that in all instances the existing shutdown margin was higher than the addition of positive reactivity caused by the moderator cooldown and under no circumstances could the reactor return to criticality after its initial scram with the existing control rod insertion limits, without any credit for operation of the charging pumps (See Reference 2, Table and Fig. 2).
To provide assurance that the reactor would remain subcritical in the post accident condition, the licensee proposed to amend the Technical Specifications by specifically requiring maintenance of sufficient shutdown margin during power operation by means of the existing control rod insertion limits and by imposing 3% margin when the reactor is in a subcritical mode.
In addition, the trip setpoint is changed to 25% rated power whenever the reactor power is below 10% of rated power.
This latter change makes the trip setpoint conform to that assumed for this reanalysis of the MSLB.
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Based on the above described staff review of the analyses we find that they provide reasonable assurance that the reactor will remain subcritical after its initial scram following a MSLB accident. We also concur with the licensee that the proposed amendments to the Technical Specifications will provide some additional assurance that the Haddam Neck Plant will be operated safely. We therefore conclude that the proposed amendment is acceptabl e.
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. 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 ES1.5(d)(4), that an environmental impact statement or negative declarc' ion and environmental impact appraisal need not be prepared in coanection with the issuance of this amendment.
Conclusion We have concluded, based upon the considerations discussed above, that:
(1) because the amendrent does notinvolve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendnent does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
i Date: October 24, 1978 1
4 References 1.
Connecticut Yankee Atomic Power Company letter to NRC (DLZiemann )
dated September 22, 1978.
2.
Connecticut Yankee Atomic Power Company letter to NRC (DLZiemann),
dated October 20, 1978.
3.
Connecticut Yankee Atomic Power Company letter to NRC (DLZiemann),
dated September 29, 1978.
4.
Connecticut Yankee Atomic Power Company letter to NRC (DLZiemann),
dated May 24, 1978.
5.
Facility Description and Safety Analysis (FDSA), Haddam Neck Plant, NY0-3250-5, Connecticut Yankee Atomic Power Company.
6.
NRC letter to Connecticut Yankee Atomic Power Company, dated May 25, 1978.
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