ML20029D042
| ML20029D042 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 04/22/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20029D039 | List: |
| References | |
| NUDOCS 9405030323 | |
| Download: ML20029D042 (9) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR R RELATED TO AMENDMENT N0. 98 TO FACILITY OPERATIN DETROIT EDISON COMPANY FERMI-2 DOCKET N0. 50-341
1.0 INTRODUCTION
By letter dated May 24, 1993, the Detroit Edison Company (Deco or the to Facility Operating License No. NPF-43 for Fermi-2. licens revises TS 3.4.3.2.d which identifies the allowable leakage for the reactorThe coolant system (RCS) pressure isolation valves listed in Table 3.4.3.2-1 b referring to the allowable values which are now listed in the table.
allowable values for the low pressure coolant injection (LPCI) loop A and B The isolation valves have changed from I gallon per minute (gpm) to 0.4 gpm for each valve.
valves have also changed from I gpm to 10 gpm for each valve Additionally, Table 3.6.3-1 has been revised to reflect that the LPCI injection line reverse flow check valves and the 1-inch bypass valves around them are no longer designated as primary containment isolation valves (
for the purpose of meeting the requirements of General Design Criteria (
54 and 55 of Appendix A to 10 CFR Part 50.
added to reflect an alternative testing method for the LPCI loop A and B isolation valves.
Type C testing requirements of Appendix J to 10 CFR Part 50 isolation valves.
2.0 QLSCUSSION On each of the two LPCI injection lines, three valves are currently specified as CIVs in TS Table 3.6.3-1, Primary Containment Isolation Valves.
operated gate valve, E11-F015, is installed outboard of containment. A motor-is remotely operated from the control room and is installed in the 24-inch Ell-F015 injection line which connects to the RCS in the reactor recirculation loop Also installed on the 24-inch injection line inside containment is a check valve, Ell-F050.
containment through the injection line. Ell-F050 is installed to prevent reverse A 1-inch bypass line is installed around E11-F050 for the purpose of system warmup for the residual heat remova (RHR)-shutdown cooling mode. A solenoid-operated globe valve Ell-F610, is normally closed except during system warmup and is remotely op,erated from the control room.
9405030323 940422 PDR ADOCK 05000341 P
The two piping penetrations are designated as primary containment pe X-13A and X-138.
"A" or "B" added to their labels; thus, valve Ell-F015A through penetration X-13A, and so on.
this discussion.
The suffixes are not used in most of -
These CIVs are subject to Type C local leak rate testing in accordance w Appendix J.
The Type C tests are performed with a gas at 56.5 psig, which is-Pa, the calculated peak containment internal pressure during a design b accident.
Coolant System Pressure Isolation Valves," as pres low pressure RHR piping.such, these valves form a boundary between As have two CIVs, either automatic or locked closed, containment, unless alternative provisions are found acceptable on some o defined basis.
valve rather than automatic or. locked closed, because L y
operate for core cooling during an accident.
The inside containment by reverse flow out of containment and Ell-F610 is locke the essential nature of the RHR-LPCI mode. automatic containme No '
v However. Ell-F050 has repeatedly failed the Type C test requiring repair and retesting.
Although the licensee has been able to repair.and successfully retest the valve, the resultant radiological and safety concerns caused the.
licensee'to review the use of Ell-F050 as a CIV under these E11-F050 is located in a hazardous area in the drywell.
within a significant radiation field.
The area is also.
outage, 10.9 person-rem were expended in maintenance and tes Li and B.
In addition, this type of work also extends the outage duration and l
increases the unavailability of one of the decay heat removal system the outage.
As a result of the review of the design basis for th containment isolation provisions of GDC 55 without'the attendant valve problems associated with the current means of_ compliance'with alternate casis examines aspects of the RHR system which had n. GDC 55.
This-in previous evaluations.
ot beenLincluded Further, the licensee has developed a basis for the AIt results
(. I Vs.
tion which justifies alternate leak rate testing provisions.ppendix J exemp 1
requirements will result in a reduction in occupational radiation exposure an
.The.new testing improvements in industrial safety while continuing to. assure containment F
integrity.
None of the valves would any longer. receive Type C tests with air" . !
as the test medium, although~ Ell-F015 would have its external leakage m (with water as the test medium) and limited to a certain value.
j 1
. 3.0 EVALUATION Two aspects of the RHR system form the basis for the proposed TS changes and exemption; (1) it is a closed system outside containment, and (') the penetrations will be water sealed during 'a loss-of-coolant accident (LOCA).
These aspects are discussed below.
3.1 Closed System Outside Containment The staff's Standard Review Plan (SRP), NUREG-0800, in Section 6.2.4,
" Containment Isolation System," provides several "other defined bases" for differing from the explicit requirements of GDC 55. Subsection II.6.e allows only a single CIV, outside containment, if the system is closed outside containment and certain other provisions are met.
The RHR system is a closed system outside containment.
It can accommodate a single active failure and still maintain containment integrity.
It is also protected against the effects of missiles and pipe whip.
The system is designed to seismic Category I standards, is classified as Quality Group B or better, and is designed to meet or exceed the maximum temperature and. pressure of the containment.
The system is also included in the American Society of Mechanical Engineers (ASME)Section XI Inservice Inspection Program and receives the required non-destructive examinations for Class 2 piping.
These programs require the system to be inspected at pressure and any visible leakage to be promptly repaired.
The RHR system is also subjected to the inspections required by the Fermi 2 system leakage reduction program. This program is a commitment made by the licensee in response to NUREG-0737, " Clarification of THI Action Plan Requirements," to conduct inspections to reduce and maintain leakage to as low as practical levels from systems outside of the primary containment that could or would contain highly radioactive fluids during or after a severe transient or accident.
The inspections are done on an 18-month frequency and the-acceptance criterion for the RHR system is 40 ml/ min external leakage for each RHR division.
The piping from the inboard check valves E11-F050A and B to valves E11-F015A and B conforms with ASME Section III, Class I requirements.
Since this piping fulfills the design requirements stipulated in Branch Technical Position HEB 3-1, no pipe breaks or cracks are postulated.
The RHR system is currently considered an acceptable closed system outside containment for the purposes of meeting GDC 56 for penetrations X-223A through D.
These penetrations are for the RHR pump suction from the suppression pool.
. The above design provisions are in accordance with SRP 6.2.4, II.6.e.
Therefore, penetrations X-13A and 8 need only have one CIV each, outside defined basis," said basis being defined by the SRP. containme The licensee proposes to no longer designate E11-F050A and B and E11-F610A B as CIVs.
high/ low pressure isolation valves in accordance with TS 4.
finds this to be acceptable, as discussed above.
The staff This will, of course, eliminate the requirement to Type C test the valves, because Type C testi applies only to CIVs.
3.2 Water Seal ClV leakage can be categorized as either through the valve seat or external to the valve (such as a stem or bonnet leak).
which show that water seals would exist at Ell-F015, the CIV outsideThe li containment, which would prevent the leakage of containment atmosphere th the valve. via either mode, during a LOCA, despite the most limiting single active failure.
The two types of water seals are discussed below.
3.2.1 Through-seat water seal Operation of the RHR pumps assures that any through-seat leakage for valve single active failure.E11-F015 will be water leakage inward towards the conta containment atmosphere through the seat and out of containment.T The following design features assure the water seal:
1.
Each division of RHR has two pumps each of which are fed from a separate diesel generator.
2.
The two divisions for RHR are cross-connected by a single header through the Ell-F010 valve which is normally kept open.
3.
Valves Ell-F015A and B, E11-F010, and the two recirculation pump valves are electrically fed from the swing bus.
The swing bus design assures power availability to each of the above valves in case one of the divisional power supplies is lost.
4 One or more of the RHR pumps will be run following an accident for either LPC1 injection or suppression pool cooling for at least a period of 30 days.
In some cases, both functions will be carried out simultaneously.
however, then RHR pumps will be operating in both divisions.Th 5.
In the long term, the RHR pumps draw suction from the suppression pool, providing a continuous or unlimited supply of water for at least 30 days.
. A detailed analysis is.provided in the licensee's submittal, but it can be seen that a water seal, pressurized to greater than 1.1 Pa (62.15 psig), would be provided at the Ell-F015 valve for at least 30 days following the onset of a LOCA, despite the most limiting single active failure, preventing through-seat leakage of containment atmosphere. However, one or both of the valves Ell-F015A and 8 may be closed during a LOCA, and this water seal might not prevent leakage external to the valve (e.g., stem or bonnet leakage); thus, the need for the following analysis.
3.2.2 External leakage water seal The configuration of the LPCI injection lines is such that a water leg would be present inboard of E11-F015 post-accident.
The licensee has provided in its submittal a conservative and very detailed analysis to show that the water leg would be at least 23.4 ft, long, in a 24-inch diameter pipe.
The staff-has reviewed this analysis and finds it to be acceptable.
The licensee proposes a limit of 5 ml/ min external leakage from Ell-F015.
It can be seen that, at that leak rate, the water leg would last more than 30 days, so that no containment atmosphere would leak out of containment during that period.
Through-seat leakage would not deplete the water leg, because, as discussed above, through-seat leakage would be water in toward containment, not out.
External leakage testing would be performed as part of the pressure isolation valve (PIV) leak rate testing required by TS at 18 month intervals, with water as the test medium and at 1045 psig.
This pressure is much greater than any accident pressure. However, this is Water is the appropriate test medium because the valve would Valve through-seat leakage is also measured by this test and limited to 0.4 gpm by the TS, but this limit is not necessary for the maintenance of either water seal.
Although the external leakage water seal would prevent atmosphere from leaking out of containment, it does not satisfy the requirements for a water seal contained in Appendix J.
from normal Type C testing with air, but it requires the water seal to b pressurized to at least 1.1 Pa during an accident. The water leg inboard of Ell-F015 does not meet this requirement, nor does it meet the spirit of the rule, wherein a water seal would at worst make water go into containment, rather than let it leak out.
Nevertheless, the staff finds that the licensee's analyses of the water seals provide sufficient assurance that containment atmosphere leakage out of containment will be prevented during an accident, to justify the granting of the requested exemption from Type C' testing of valves Ell-F015A and B with air as the test medium.
In lieu of that, the licensee will measure external valve leakage in conjunction with PIV leak rate testing and limit it to 5 ml/ min.
. 3.3 Pressure Isolation Valve Allowable Leakaae Chances which also function during RHR.The subject valves are a pair of val Anchor Darling swing check valve (E11-050 A/BThe inboard valve of each pair n
outboard valve of each pair is a 24" motor-ope) rated flexible wedgin (Ell-F015 A/B) outside containment.
e gate valve leakage valves for the pressure isolation function from theThe TS proposes I gpm allowable leakage, to 10 gpm and 0.4 gpm, for the check v l current maximum of valves, respectively.
a ves and gate The current Fermi-2 TS require PIV leakage to be li gpm. This limit the limits were changed in Standard Technical Specificati at which time nominal inch of valve size, up to 5 gpm.
ons to 0.5 gpm per establish allowable leakage rates for the " Event V" d ment to The " Event V" scenario was the failure of two check valves n WASH-1400.
subjecting a low pressure system outs n series containment to full reactor pressure,ide of a pressurized-water reactor (PWR) causing a LOCA which bypassed containment. rupturing the low pressure piping and V" valves was 1.0 gpm, with leakage between 1.0 an The leakage criterion for " Event the measured leakage rate and the maximum permissible (5 e margin between greater.
gpm) by 50% or For subsequent near-term operating license requirement for inservice leak testing was e(xtended to all PIVNT0L) applicatio second valve in series leading away from reactor coolant pressure ()irs s f inside or outside containment and including both boiling water either plants.
criterion for " Event V" valves was imposed on all PIVsIn gen reactor and PWR maximum leak rate of 1 gpm, generall For NTOL reviews, a all PIVs.
i The stricter acceptance c/ without qualification, was imposed for and in better condition.because it was considered that the valves could me e standard being newer the capacity of the pressure relief systems in most of thThe limit was e plants.
The staff later determined that the 1 gpm acceptance criteri l
indicator of imminent accelerated deterioration of valves o i
on was not an valve failure.
Regardless of the leak rate allowed for each PIV potential limit the allowable leakage from the RCS as a whole
, plant TS number of valves in the RCS. effectively eliminates the possibility of large i
therefore, have no impact on the total allowable RCS le kIncreasing the age from a a age.
The NRC contracted a study of PIV leak test requirements (EGG NTA
" Inservice Leak Testing of Primary Pressure Isolation Valves " F b P-6175, which presented an assessment of PlV inservice leak testing a, d leakage limits.
e ruary 1983)
The report recommended that the owner be allowed the optio n allowable of a higher allowance of leakage for specific valves if justifi d b analysie of overpressure protection and radiological processing capab n
e y an
_ _ _ _ _ - _ - - - - - - - - - ~ ' " ' ^ ~ ' ' ~
, showing that the ASME Boller and Pressure Vessel Code (the Code), Sectio and the plant safet leakage allowance. y analysis conclusions are not violated by the higher-The primary objective of the staff in allowing higher leak' rates was to decreaseL the time spent on unnecessary maintenance on the valves ?
which attributes to faster deterioration of the valves and increased exposure -
to personnel.
The choice of-an absolute maximum allowable leakage rate of 5 gpm was based on experience gained from the plant using 5 gpm to comply with 0
" Event V" orders.
Allowable leak rates above 5 gpm were not considered conservative because of a lack of experience.
The EG&G report identified the following reasons for_ limiting the leakage'of reactor coolant into lower pressure interfacing systems, such as the LPCI system:
('1) Leakage from the RCS, together with other sources may exceed the flow capacity of the pressure relief system, causing overpressurization of.a lower pressure system.
(2) A large allowance of identified leakage through primary pressure isolation valves may make it more difficult for leakage detection systems to identify small but important increases in unidentified leakage.
(3) Leakage of radiologically contaminated RCS water may. exceed the proce capacity of waste water cleanup systems.
(4) This allowed breach of containment may increase the probability of uncontrolled fission product release under certain accident. conditions.
EGG-NTAp-6175 concluded that the TS limiting values for leaka more than 5 gpm leakage was not recommended because or 0.5 gpm per inch of diameter up to 5 gpm, are acce An' allowance of accounted for in the ASME Bof1er-and Pressure Vessel Code 1) the leakage may not-be1 Section III overpressure protection analysis, (2) the larger leakage a,llowances wou,ld tend.
to mask the detection of unidentified leakage from the pr to limit fission product distribution from the primary coolant to plantf systems outside the containment.
the two in-series val _ves is maintained below the previous. total allow gpm versus 1 gpm), which results in a more conservative limit for. all cas(0.4 except-inadvertent opening of the motor-operated gate valve (which is'-
es interlocked closed below shutdown primary system pressure maximum leakage of 10. gpm, each of these three-factors are).
Even assuming the considered'and=
discussed below in relation to the proposed TS change:
(1) The analysis prepared by Deco addresses the overpressure protection analysis of the RHR system, indicating that a 1-inch relief nlve on each of the low pressure side of.the pressure isolation. valves has~ a capacity.
of 290 gpm at 450 psig.
The' Fermi-2 updated final safety analysis re (UFSAR), Section 5.5.7.3.5, states that the " reactor coolant system port pressure boundary isolation valve leakage is accommodated by 1-in. relief valves.
This size of the valve is considered large enough to' accommodate R
any postulated-leakage."
The analysis indicates that the UFSAR evaluation y
I mr 14,u-m
s-t 8-is based on a PIV leakage test acceptance criteria of 10 gpm. Therefore a leakage limit of 10 gpm appears to be acceptable from an overpressure,
protection concern.
(2) The_ analysis submitted by Deco did not specifically address _the i
possibility of masking the detection of unidentified leakage from the primary system due to an increase of the allowable leakage limit of the check valves; however, by specifying a lower limit for the gate valve (0.4 gpm), which is interlocked closed with RCS pressure'above approximately L
450 psig, the pair of valves will maintain leakage below a level of concern for increasing the possibility of masking leakage. The total unidentified RCS leakage allowed by TS is a conservative 5 gpm.
Therefore, an increase in the leakage limit for the check valve does not create a safety concern.
(3) For concerns of limiting leakage outside containment due to PIV leakage, the Deco analysis for the TS change discusses this aspect in terms of containment isolation issues; however, the results are applicable to PIV leakage as well.
The two series isolation valves provide redundancy for each of the two RHR injection lines. The motor-operated gate valves are interlocked with a' pressure switch that prohibits opening of the valve if the recirculation pressure exceeds the shutdown range (Fermi 2 UFSAR Section 5.5.7.3.5).
As noted in item (1) above, the relief RHR valve is of a size 19 relieve the maximum allowable leakage af either valve (10 gpm) with the discharge to inside containment.
The RHR system forms a closed system outside containment, with a water seal maintained as a result of the physical configuration of the LPCI loop selection feature of the RHR system.
The RHR system is protected against the. effects'of
't missiles and pipe whip, is subject to inservice inspection for ASME' Class 2 piping, and is included in the leakage reduction program established in accordance with NUREG-0737 to reduce leakage of primary water outside cont ainment.
The piping between the inboard check valve.and the outboard motor-operated gate valve conforms to ASME Class I piping requirements.
The combination of these design features and monitoring programs serve to limit leakage outside containment; therefore, the increase in allowable leakage will not create a safety concern.
Based on the foregoing evaluation, the staff finds that the licensee has provided adequate bases for changing the allowable leakage values for RCS pressure isolation valves listed in Table ?.4~.3.2-1, specifically the:LPCI loop A and B isolation valves and the loop A and B testable check valves, from 1 gpm to 0.4 gpm and 10 gpm, respectively. Additionally, the staff finds that the licensee has provided adequate alternative bases for' meeting GDC 54 and 55 and that the LPCI reverse flow check valves and 1-inch bypass valves may be removed from the list of primary containment j
0 isolation valves in 3.6.3-1.
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4.0 STATE CONSULTATION
'1 In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment.
The State official had no comments, l
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5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes a surveillance requirement The staff has determined that the change in the types, of any effluents that may be released o there is no significant increase in individual or cumulative occupa,tional and that radiation exposure.
The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (58 FR 46227).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.32, an environmental assessment of the exemption from certain requirements of 10 CFR Part 50, Appendix J, related to these actions was published in the Federal _ Reaister on April 21, 1994 (59 FR 19028).
will not result in any environmental impacts beyond those Fermi-2's final Environmental Statement.
6.0 CONCLUSION
The staff has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safet will not be endangered by operation in the proposed manner,y of the public (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the comm defense and security or to the health and safety of the public.
Principal Contributors:
J. Pulsipher P. Campbell Date: April 22,1994
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