ML20211N176
| ML20211N176 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/20/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211N162 | List: |
| References | |
| NUDOCS 8703020009 | |
| Download: ML20211N176 (4) | |
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UNITED STATES g
NUCLEAR REGULATORY COMMISSION n
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WASHINGTON, D, C. 20555 1
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR PEACTOR REGULATION SUPPORTING AMENDMENT NO.101 TO FACILITY OPERATING LICENSE NO. DPR-29 AND AMENDMENT N0. 98 TO FACILITY OPERATING LICENSE NO. DPR-30 COMMONWEALTH EDIS0N COMPANY AND IOWA-ILLIN0IS GAS INU ELECTRIC COMPANY b
QUADCITIESNUCLEARPOWEUTATION, UNITS 1AND2 DOCKET NOS. 50-254/265
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1.0 INTRODUCTION
By letter dated November 27, 1984 as supplemanted July 22, 1986, Commonwealth Edison Company (CECO, the licensee) proposed amendments to Appendix A, the Technical Specifications (TS), for Operating License Nos. DPR-29 and DPR-30.
The proposed changes would raise the drywell high pressure trip setpoint from 2.0 psig to 2.5 psig, delete the existino hiweekly main steam isolation valve surveillance, and delete a note on page 3.7/4.7-10 in Appendix A to DPR-29. The July 22, 1986 supplemental submittal provided new proposed TS pages as a number of administrative typographical errors were identified in the November 27, 1984 submittal.
This Safety Evaluation is a review of the requested changes and their impact on the operation and administration of plant activities.
2.0
SUMMARY
OF EVALUATION The changes proposed by the licensee to the high drywell pressure trip setpoint will reduce the frequency of spurious Engineered Safeguards Features (ESF) actuation without measurably impacting existing safety margins. The change proposed to the frequency of Main Steam Isolation Valve (MSIV) testing is reflective of the demonstrated reliability of the MSIVs and is consistent with Standard Technical Specifications (STS) requirements. Deletion of the footnote on page 3.7/4.7-10 of the Unit 1 TS is reflective of restored eauipment operability.
i The staff agrees with the changes as described in the proposed t.mendment.
0703020009 870220 3.0 EVALUATION PDR ADOCK 05000254 P
PDR 3.1 Proposed Changes to Unit 1 and Unit 2 Technical Specification Tables 3.1-1, 2, 3 and 3.2.1, 2 These tables are revised to reflect the increase in the high drywell pressure trip setpoint from 2.0 to 2.5 psig. The high drywell pressure trip signal is used to initiate certain ESF systems in response to a loss-of-coolant accident (LOCA).
An increase in the high drywell pressure trip setpoint from 2.0 to 2.5 psig must be evaluated for its effect on the
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. safety systems to which it provides an actuation signal and the potential for an increase in peak containment accident pressure.
An increase in the high drywell pressure trip setpoint could impact plant response to accident conditions in the following ways:
An increased trip setpoint would allow plant operation with a higher a.
nominal drywell pressure. Hence, initial pressure at the beginning of an accident could be higher. This impacts directly the drywell pressure throughout the transient, the volume of water in the torus, primary system blowdown rate, and Emergency Core Cooling System (ECCS)
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flow rates.
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b.
The time required to reach the trip setpoint during an accident will be longer at the higher trip point. Consecuently, ESF actuation in response to high drywell pressure may be delayed.
With respect to Item a above, the licensee currently controls drywell pressure at a nominal 1.3 psto by approved procedures. As confirmed by discussions with the licensee, no change has been made or is contemplated 4
to exit ting procedurally controlled drywell pressure limits.
Howe'ver, if drywell pressure were to be maintained at an elevated value, the effect on plant transient behavior would be within analyzed limits for the following reasons:
A post accident peak drywell pressure of 47 psig has been calculated a.
fo' the Quad Cities drywell. This value assumes that the drywell is at 3tmospheric pressure prior to the accident. Using the conservative assumption that the peak drywell pressure value of 47 psig is a pressure rise due to accident conditions, with a 2.5 psig initial drywell pressure the peak value achieved would be approximately 50 psig, well below the 57 psig drywell design pressure. Thus, drywell integrity will not be compromised.
b.
Technical Specifications require that a specified minimum volume of water be maintained in the torus to satisfy energy absorption requirements.
If drywell nominal pressure is increased, water will be forced from the drywell through torus vents into the torus. This will, in turn, cause indicated water level in the torus to increase.
Operator action to restore indicated torus water level would reduce the torus water inventory. This effect has been analyzed and procedural limits are in place to maintain sufficient water in the torus to satisfy energy absorbing requirements.
c.
An increase in drywell nominal pressure at the time of the accident would not appreciably affect primary system blowdown rates as the blowdown occurs under choked flow conditions where flow is essentially independent of downstream pressure.
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ECCS pump performance would be essentially unchanged in the case of a slightly elevated drywell pressure for the following reasons:
(1) Reactor coolant system pressure versus' time is essentially unchanged by a small increase in drywell pressure owing to the choked conditions of break flow. Hence, ECCS pump discharge 4
pressure requirements for those pumps injecting into the reactor coolant system remain unchanged. Pump capabilities would, in fact, be slightly improved owing to the increase in available suction head resulting from increased drywell pressure.
4 (2) Those pumps discharging directly into the drywell free air space (containment spray pumps) would experience an increased resist 3nce to flow due to higher drywell pressure. This would be partially offset by an increase in available pump suction head resulting from increased drywell pressure. However, as noted in Figure 5.2.17 of the Quad Cities Final Safety Analysis Report, containment pressure remains below design values even in the absence of containment spray flow.
With respect to Item b above, the time required to achieve high drywell pressure ESF actuation is a function of the difference in pressure between i
drywell ambient and the trip setting. The original Quad Cities analysis assumed that this difference was 2 psig based on the drywell being at atmospheric pressure (0 psig). Mark I Containment studies and resultant analysis have resulted in the current Technical Specification (TS) i requirement of normal minimum drywell pressurization of 1.?O psig above the containment wetwell which is normally atmospheric (0.0 psig) or 1.20 psid.
As stated above the original Quad Cities analysis assumed a difference of 2.0 psig between the ESF actuation trip setting and drywell ambient. Also, as stated above, the current TS require nomal drywell minimum pressure of l
1.20 psig. Therefore, since the current trip setting is 2.0 psig, the i
current actual pressure differential to actuation is only 0.8 psig (2.0 psig trip - 1.2 psig ambient) even though the current analysis supports a differential as large as 2.0 psig. The time to ESF actuation is larger for t
I 2.0 psig and, therefore, bounding.
Therefore, i# the trip setting is increase to 2.5 psig, assuming no change in drywell pressure control requirements, the differential pressure between l
the new trip setting and drywell ambient would be 1.3 psig (2.5 psig - 1.2 psig ambient = 1.3 psigl.
Since 1.3 psig is less than 2.0 psig, the i
original analysis bounds this change; therefore, the staff finds this i
change acceptable.
l 3.2 Proposed Changes to Unit I and Unit 2 Technical Specifications Surveillance Requirements 4.7.D.I.d and Surveillance Requirement l
Basis 4.7.D l
l This surveillance requirement and basis are revised to reflect the deletion of the biweekly Main Steam Isolation Valve surveillance. The purpose of the biweekly MSIV surveillance test was to ensure that the MSIVs would
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close when required to do so. This test was a result of several instances in early 1971 where one or more MSIVs did not isolate when required.
Continued testing has demonstrated the MSIV closure function to be highly reliable.
The General Electric STS, NUREG-0123, and Quad Cities Technical Specifications require certain containment isolation valves, including MSIVs, be tested for operability on a quarterly basis. The STS do not require that the MSIVs undergo a biweekly partial closure test. Thus, the proposed change is consistent with STS requirements.
On these. bases, the deletion of the biweekly MSIV surveillance test and i
corresponding change to the TS basis are acceptable to the staff and do not adversely affect the safety of the plant or the health and safety of the j
general public.
3.3 Proposed Change to Unit 1 Technical Specifications 3.7/4.7 Page 10 i
This Technical Specification change deletes an obsolete note at the bottom of the page which required valves M01-220-2, 3 and 4 be closed during i
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Operating Cycle 7 or until valve M0-220-1 was restored to operable status.
Since valve MO-220-1 is now operable, this change does not adversely affect the safety of the plant or the health and safety of the public, and is j
acceptable to the staff, i
4.0 ENVIRONMENTAL CONSIDERATION
S j
The amendments involve a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes in surveillance requirements. The staff has determined that i
the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released j
l offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards l
consideration and there has been no public coment on such finding.
Accordingly, the an:endments rieet the eligibility criteria for categorical i
exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no I
j environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments, i
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5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public j
will not be endangered by operation in the proposed manner, and (2) such j
activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the common
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defense and security or to the health and safety of the public.
l Principal Contributors:
S. Hare, W. G. Guldemond, T. Rotella j
Dated:
February 20, 1987 I
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