ML20153G185
| ML20153G185 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/29/1988 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20153G190 | List: |
| References | |
| B12132, NUDOCS 8805110164 | |
| Download: ML20153G185 (29) | |
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(203) 665-5000 April 29, 1988 Docket No. 50-245 B12132 Re:
10CFR50, Appendix J U. S. Nuclear Pegulatory Commission Attn:
Document Control Desk Washington, D. C.
20555 Gentlemen:
Millstone Nuclear Power Station, Unit No. I Integrated Safety Assessment Program Topic Nos. 1.14, "Appendix J Modifications," and 2.33. "RBCCW Leak Rate Testina" In a letter dated November 19, 1986,II) Northeast Nuclear Energy Company (NNECO) provided the status of Millstone Unit No l's request for exemptions from the requirements of 10CFR50, Appendix J.
Transmitted herewith is NNECO's request for exemptions relating to the Type C testing requirements of 10CFR50, Appendix J, Section III.C with respect to certain containment penetrations.
The exemption requests are provided in Attachment 1, together with the techni-cal justification for each exemption.
For those exemptions analyzed by a Probabilistic Risk Assessment (PRA), the PRA calculations are provided in.
The list of penetrations for which exemptions are being sought is not the same as the list of potential exemptions in the November 19, 1986 letter.
After further analysis under ISAP Topic Nos.1.14 and 2.33, using the ISAP Public Safety Impact Assessment methodology, it was determined that an exemption will not be sought for some of the penetrations identified in the November 19, 1986 letter.
The Commission's regulations, specifically 10CFR50.12(a), provide that exeap-tions may be granted from the requirements in 10CFR50 if "special circumstanc-es" are present and the exemptions are "authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security."
Under 10CFR50.12(a)(2)(ii), "special circumstances" warranting an exemption are present whenever, among other things, application (1)
Letter from J. F. Opeka to C. I. Grimes, "Supplement to Integrated Safety Assessment Program Topic No. 1.14, ' Appendix J Modifications'," dated Aoi November 19, 1986.
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2 PDR ADOCK 05000245 l/
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U.S. Nuclear Regulatory Commission B12132/Page 2 April 29, 1988 of the regulation in the particular circumstances "is not necessary to achieve the underlying purpose of the rule..."
On the basis of the information provided herewith, NNECO concludes that the exemptions for the subject penetrations are justified under the standards of 10CFR50.12. As demonstrated in Attachments 1 and 2, the requested exemptions will not present an undue risk to public health and safety since no signifi-cant increase in the risk associated with containment leakage will result from a grant of the exemptions.
In fact, granting the exemptions will result in a reduction in occupational exposure.
Many of the modifications that would be required to satisfy the specific requirements of Appendix J would have to be made in the drywell area and would result in more worker exposure than could be justified.
Furthermore, "special circumstances" are present in that application of the regulation is not necessary to achieve the underlying purpose of Appendix J.
The purpose of the leak rate testing requirements in Appendix J is to provide reasonable assurance that containment leakage during a design basis event will not exceed the leakage limits specified in binding licensing documents or assumed in the pertinent accident analyses.
The Type C testing requirements of Section III.C of Appendix J are established to provide adequate assurance that the leakage rate from containment penetrations is within acceptable limits. As shown in Attachments 1 and 2, the requested exemptions from Type C testing requirements for the selected penetrations will satisfy the underlying purpose of the rule since containment leakage will be held within the allow-able limits..Given the system configurations and otner factors presented in the exemption request, there is adequate assurance that the penetrations in question will remain sufficiently leak tight.
This will be confirmed by the Integrated Leak Rate Test (ILRT), which will ensure that the overall contain-ment leakage rate is sufficiently limited.
NNEC0 has determined that modifications to permit Type C testing of valves IC-6 and IC-7, which are in the isolation condenser vent piping to the main steam system (penetration X-10A),
are warranted.
These 3/4" solenoid /
diaphragm actuated control valves are normally open to vent noncondensible l
gases from the isolation condenser to the "A" main steam line.
They are automatically closed on a Group 1 isolation, a Group 4 isolation, and whenever valve IC-3 is opened (i.e., upon isolation condenser initiation).
The isola-tion condenser is considered to be a closed loop outside of containment.
As such, valves IC-6 and IC-7 become boundaries of this closed loop during operation of the isolation condenser.
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Since the isolation condenser is open to the vented reactor coolant system during the ILRT, these valves are periodically leak checked at 43 psig.
Due to excessive leakage through penetration X-10A during the Type A test at Millstone Unit No. I during the 1987 refueling outage, NNECO currently plans 1
to implement modifications which will permit Type C testing of IC-6 and IC-7 i
during the 1989 refueling outage.
These plans will be finalized during the development of our next Integrated Implementation Schedule.
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812132/Page 3 April 29, 1988 NNEC0 has also determined that modifications to permit Type C testing of drywell sump drain valves SS-3, SS-4, SS-13, and SS-14, contained in penetra-tions X-18 and X-19, are warranted.
This conclusion was reached even though these penetrations are associated with sumps, and are generally water sealed.
In a design basis accident, these sumps will receive reactor coolant from the assumed recirculation line break.
This water will act as a seal between the containment atmosphere and the containment isolation valves.
It is expected that, if these valves were to leak following a design basis accident, substan-tial dilution, cooling, and iodine removal would occur prior to any potential j
release.
In addition, subject valves are exposed to 43 psig contatu.1ent atmosphere during ILRTs (the sumps are pumped down for this purpose).
However, the public safety benefit (potential radiological exposure) to be gained by leak testing these valves, as determined from our PRA calculations, was high enough to warrant modifications to permit Type C testing.
As such, NNECO commits to implement modifications to penetrations X-18 and X-19, on a schedule to be determined after inclusion of this project into the Millstone Unit No.1 Integrated Implementation Schedule.
Although complinice with 10CFR50, Appendix A is beyond the scope of this submittal, the interrelationship of the criteria of Appendices A and J neces-sitates discussions concerning the provisions of Appendix A in our exemption request for the containment isolation valves in the reactor building closed cooling water system.
NNECO is currently developing a submittal which will provide additional information to the NRC Staff regarding containment isola-tion valves and our inservice testing program at Millstone Unit No.1.
This submittal will the issues raised in Inspection Report No. 50-245/88-02.(2jiddress Inherent in our submittal will be discussions of certain provisions of 10CFR50, Appendix A.
In summary, NNECO has concluded that exemptions for those areas discussed in Attachment I are warranted under the standards of 10CFR50.12.
It should also be noted that NNEC0, through ISAP Topic 1.14,. has evaluated a number of plant modifications to achieve compliance with Appendix J requirements.
The installed and planned modifications represent prudent steps to improve the containment integrity of Millstone IJnit No. I and demonstrate NNEC0's good faith efforts to satisfy Appendix J.
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Bettenhausen letter to E.
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- Mroczka, "Resident Inspection 50 245/88 02," dated March 4, 1988.
l 4-U.S. Nuclear Regulatory Commission B12132/Page 4 April 29, 1988 In accordance with 10CFR170.12(c), enclosed with this exemption request is the application fee of $150.00.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY l'
'e Senio/MFoczka //
E. 'U r Vice President cc:
W. T. Ru:ssell, Region I Administrator M. L. Boyle, NRC Project Manager, Millstone Unit No. 1 W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3.
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Docket No. 50-245
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Millstone Unit No.'1.
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. Requests for Exe.nption from 10CFR50, Appendix J Requirements
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B12132/Page 1 PI. ANT:
MillstoneUnitNE.1 l
j TITLE:
Exemption from 10CFR50, Appendix J,.in Penetration X-20 l
DESCRIPTION WECO requests exemption
- rom Type C tesking of containment isolati6n j
barriers in this dominerali::nd water system penetration. N isclation barr.iers_ in this penetration are EM-64, -66, and -67.
1 JUSTIFICATION BASES ihe 1-inch manual. containment boundary valves in this penetration are normally locked closed and administratively controlled. 'Ihese valves 3
remain closed during and after a design basis accident. Beso conditions put these small lines into the same category as small test branches off of conteinment.
Functional leak checks.of these branches into contain-I ment, performed after each use, are proposed as a means of satisfying Appendix J.
Functional testing would be considered acceptable with zero leakage into containment or out of piping through containment when these valves are shut after use..
In addition, this small branch within containment is subjected to leakage testing during Appendix J Type A testing (i.e., it is a containment
-boundary within the sec>pe of the Type A test).
i mECO believes that an exemption from Type C testing of the subject valves is justified on these bases.
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i 812132/PVA 2 PIANT:
Millstone Unit No. 1 I
TITLE:
Exemption from 10CrR50, Appendix J, in Penetration X-21 DESCRIPTION 4
M4ECO requests exeaption from Type C testing of containment isolation valves (check valve SA-345 and motor-operated gate valve SA-344) in this service air penetration.
JUSTIFICATION BASES i^
1he service air piping through penetration X-21 is 1" piping that ends in normally closed manual valves at each air hose connection point.
In, l
addition to these boundaries, containment valve SA-344 is normally closed and administratively controlled by system operating procedures. There-fore, this penetration ems in a closed piping branch within containment.
The piping network within containment is subjected to leakage testing'as a boundary during Appendix J Type A tests. Also, it is possible that 100 psig air pressure in Penetration X-21 will'be available to act as a gas seal at the penetration (i.e., a gas charge can be trapped between the containment valves and the dead-ended hose branches within the drywell
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when SA-344 is closed after use)..
Due to the multiplicity of berriers described above (three valves in I
series and a potential gas pressure seal) an exemption from Type C testing is requested.
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i B12132/Page 3 PLART:
Millstone Unit No. 1 I
TITLE:
Exemption from 10CFR50, Appendix J in Penetration X-42 DESCRIPri.ON NNECO requests exemption from Type C testing of check valve SL-8 inside the drywell in the standby liquid control (SLC) system.
JUSTIFICATION BASES Presently, there are no test connections to permit a Type C test to be performed for SL-8.
Modifications for teating and subsequent testing will result in more personnel exposure than is desirable. W outside
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check valve in'this penetration (SL-7) currently undergoes Type C testing.
W SLC system is a safety-related system which is isolated from Penetration X-42 until explosive ("squib") valves (SL-5A and -5B) are fired. Diaphragms in these valves act as passive barriers in this penetration.
Firing the squibs rhears these diaphrages and permits bocated water to be injected directly into the reactor vessel. When the i
SLC system is injecting borated water to the rector vessel at approx-imately 1250 psig, leakage out of containment into this system is not possible.
It is important to note that the firing of squib valves and i
actuation of the SLC system are not actions considered necessary for recovery from a loss of coolant accident.
l h leak tightness of the SLC system is functionally demonstrated during testing. On a monthly schedule, the system is tested in the recircu-lation mode (pumping water to the test tank from the SLC system tank).
Less frequently (once per core cycle; performed during refueling), an explosive valve is fired and high pressure domineralized water is in-1 jected into the reactor vessel.
h SLC system is Category I.
It has been designed in accordance with the Power Piping Code (B31.1) and Seismic Category I criteria. W re-fore, it can be assumed this system, through penetrations X-42, will remain intact and functional in a design basis accident (DBA).
Since the SLC system is isolated from X-42 in a DBA by closed piping and a leaked tested containment valve (SL-7), it is considered unnecessary to rely on SL-8 as a containment barrier.
Exemption from Type C testing of SL-8 is requested on this bt41s.
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~ Attachment 1 B12132/Page 4 PLANT:
Millstone Unit No. 1 TITLE:
Exemption from 10CFR50, Appendix J in Penetration X-22 DESCKIPTICN
!NECO requests exemption from Type C testing of the containment isolation valves (AC-162 and AC-50) in this drywell compressor discharge per.etra-tion.
JUSTIFICATION BASES he piping in this penetration has two branches. Each branch has additicnal seismic, m Category I, check valves, which can provide an additional leakage barrier (these valves are AC-49 und AC-51). here-fore, there are three barriers to release in each path out of this pene-tration.
Also, the drywell compressor generally provides 95 psig gas to the containment through this penetration.
If this penetration is closed in a design basis accident (DBA) and the drywell conpressor system remains intact (it is not seismic), residual pressure in the piping downstream of the containment valves (AC-162 and AC-50) will provide a gas seal at this penetration. This gas seal will be at a pressure greater than post-DBA 1
containment pressure and, therefore, will prevent leakage out of this penetration.
We secondary N,d capable of charging 95 psig gas to Penetration X-22 gas supply system is seismic and % Category I.
It will i
remain intact an after simultaneous occurrence of a DBA, a safe shutdown earthquake and a i
loss of normal power.
For reasons very similar to those given above, it is very likely that a 95 psig gas charge will be tespped between the N nupply system and X-22 in a design basis accident if and when valve AC'-50 is shut (AC-50 is not automatically shut in a DBA). his gas seal will prevent leakage out of the subjer.t penetrations.
Based on the multiplicity of barricts (three valves and a potential gas pressure seal), an exemption from Type C testing is requested.
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s B12132/Page S PIANT:
Millstone Unit No. 1 TITLE:
Exemption from 10CFR50, Appendix J in Penetration X-212 DESCRIPTION NNECO requests exemption from:
Appendix J, Type C, leak testing requirements for valves in this penetration and all of its branches. Penetration X-212 is the reactor water cleanup (MCU) system vent line routed into the torus.
JUSTIFICATION BASES ne justification for exemption from Appendix J requirements is based on the premise that the 4' torus Vater seal at Penetration X-212 is undeple-table. NNECO considers the water seal at penetration X-212 to be a con-tainment isolation boundary. his boundary supplements the following additional boundaries provided in each branch off of X-212.
Ch :k valve CU-69 is considered a containment isolation valve for o
alt branches.
Relief valve CU-68 is considered a containment isolation valve for o
the RWCU system vent branch. Relief valves are considered accept-able containment boundaries per ANSI N271-76/ANS 56.2.
h e isolation condenser trap drain line is Seismic Category I.
It o
is considered a closed loop routed back to the isolation condenser syr ?.em.
nis closed loop acts as a barrier to containment reltases from X-212.
o Normally closed valve (CU-71) in the snall drain line routed to the reactor building equipment drain tank acts ns a containment bart:!er.
We 4' water seal ovar the opening of the X-212 pipe in the torus is equivalent to at least 240,MO gallons of water. We leak tightness of the torus and this water seal are demonstrated on a continuous basis, since the torus level is monitored at least once per shift during plant operation using instrumentatic" in the control room.
De LPCI and CS branches off of the torus are hnetionally tested each month. Also, the torus water level can be replenir>hed from external sources if dose rates in the torus room permit personnel entry.
Finally, it is important to note that th9 containment boundaries at X-212 are subjected to leakage checking during Appendix J Type A tests.
On these bases, NNECO believes that an exemption from Type C testing in this penetration is warranted.
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1 PLANT:
Millstone Unit No. 1-TITLE:
Exemption from Reverse Direction Testing Requirements cf I
10CFR50 Appendix J in Containment Atmosphere Control Butterfly valves
REFERENCE:
U.S. NRC IE#Informaticn Notice No. 86-16 entitled "Failures to Identify Centainment Leakage Due to Inadequate Local
'Itsting of BWR Vacuum Relief ^ System Valves" i
DESCRIPTION Section III.C.1 of Appendix'J to 10CFR50 states:
l "Type C tests shall be performed by local pressurization. %e pressure shall be applied in the same direction as that when the valve would be required to perform its safety function, unless_it can be determined that the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results" l
tEECO requests an exemption from the above requirement so that certain valves in the following penetrations may be tested in the reverse i
direction:
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o X-25 o X-26 o X-202E o X-205 BACKGROUND a
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Millstone Unit No. 1 employs many Allis-Chalmers butterfly valves as containment isolation valves in its atmospheric control (AC) system.
%ese valves are AC-3A, AC-3B, AC-4, AC-5, AC-7, AC-8, AC-10, AC-1.1, and AC-17.
Because of the orientation of test connections between these valves, they 1
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are local leak rate tested in groups by pressurizing the piping between valves. For this reason, valves AC-3A and B, AC-5, AC-6, AC-7, and AC-ll are Type C tested with pressure applied in a direction which is the re-J verse of that associated with a DBA. Valves AC-5, AC-6 and AC-7 were i
turned around to assure that revu se direction testing would be performed
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in a conservative mar;ner. Reverse direction testing is considered to be conservative if test pressure tends to push a valve disc off of its seat
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and DBA pressure tends to push the valve disc into its seat.
We subject valves have double ring seals on both mounting faces and sealed actuator shaft penetrations of their valve bodies. Re-orientation I
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of valves to accomplish conservative reverse direction testing moved the i
actuator shaft seals out of the local leak rate test (LLRT) boundary.
1 We ring seals on the containment sides of the reverse direction tested valves have always been outside of the Type C boundary. Were are test plugs on the valve bodies to permit local leak rate testing of the ring i
seals. Millstone Unit No.1 can revise its 11RT procedures for AC but-terfly valves to include a test of the ring seals when these potential-j containment boundaries are not part of a valve's Type C tests. Ring seal leakage can be added to valve seat' leakage for comparison to 11RT accep-tance criteria.
JUSTIFICATION It is not possible to pressurize the subject reverse direction tested butterfly valves from their containment sides without pressurizing the' j
entire containment. Wis is because these AC system lines end in open pipes ir: side the.:ontainment. Modifications to permit isolation of these lines are inadvisable due to space problems (with regard to new test i
boundary valves and test branch-connections) and/or the difficulty that i
would be encountered in installing test blanks at each penetration during every outage.. We penetrations range in siza;from 18" pipe to 24" pipe.
%e X-202 and X-205 penetrations 'are in the torus. Movement and instal-lation of large blanks in this structure would be troublesome (e.g.,:
1 working over water from a catwalk). In addition, the X-202 penetrations would have to be isolated from the drywell in their torus to drywell vacuum breaker branches. % is appears to require installation of fiva 18" pipe blanks in the 6'9" diameter pipes connecting the drywell to the torus. Wese modifications would be very difficult to perform, if they could be performed at all.
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2e effect of system pressure on closing force and sealing can be 4
accounted for in a valve's design. If the closing compressive force is large enough, sealing will be achieved regardless of the additive or subtractive effects of system pressure. MP-1 has leak tested atmospheric control system butterfly containment valves of this type with.the same disc / seat to containment orientation la both directions with good results (provided the soft seats are properly maintained).
For example:
AC-4, AC-8, AC-10 and AC-17 are leak tested with pressure applied in accident direction; AC-ll, AC-3A and AC-3B are reverse direction tested, i
_CONCLUSICNS i
We' objectives of equivalent or conservative reverse direction testing j
and actuator shaft seal testing can not be achieved simultaneously with existing piping configurations.
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L B12132/Page 8-WECO proposes to turn valves AC-5, AC-6,,and AC-7 around again. INCO also proposes to incorporate applicable double ring seal leakage tests into the local leak rate test procedures of affected valves. %ese steps would bring actuator shaft seals and double ring seals within the con-tainment boundaries into the Appendix J test program. his proposal i
would be the most conservative test configuration possible at this time.
To support trie proposals made above, WECO requests exengtion from Appendix J restrictions on reverse direction testing of the subject butterfly valves.
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1 B12132/Page 9 PLANT:
' Millstone Unit One i
TITLE:
Exemption from 10CTR50, Appendix J, in Penetrations X-23 and X-24 DESCRIPTION NVECO requests exemption from:
1.0 Type C testing of existing containment isolation valves in these R3CCW penetrations, a
JUSTIFICATION BASES 1.
A probabilistic risk assessment (PRA) of the subject penetrations has been performed. his assessment disclosed that the con-servatively calculated public safety benefit to be gained by modi-fying these penetrations to permit App. J testing wcald be 0.98-2.49 man-rem (for X-23 and X-24 respectively). %ese potential public exposure savings would be reduced to O man-rem if proposed seismic upgrades to the RBCCW system inside containment are completed. An estimation of the modifications required to comply with Appendix A General Design criteria and to accomplish Appendix J testing disclosed that these plant changes could require as little as 2.7 man-rem (if only two new containment valves are required in tne reactor building) and as much as 13 man-rem (if four 'new containment valves are required, two in the drywell and two in the reactor building). The cost of these modifications would range between i
$840,000 and $1,256,000. These cost and man-rem estimates do not include any seismic upgrading work on the RBCCW loop inside the containment.
l 2.
We RBCCW system is considered a closed loop inside and outside
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containment for the following reasons (Reference ANSI N271/ANS 56.2-76 and NUREG-0800, Section 6.2.4.II.6,0):
o RBCCW does not connunicate with reactor coolant or the containment atmosphere.
he RBCCW system demonstrates its leak tightness functionally o
during its continuous operation while the plant'is in the "run"
- node, he RBCCW is designed to be functional for recovery from plant i
o incidents at the operator's discretion (e.g., the system requires remote manual actuation to accomplish containment s_
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B12132/Page 10 isolation; during loss of normal power' incidents, the RBCCW k
i pumps are automatically aligned to emergency _ power for possible use). his implies that the system is expected to remain 4
intact at design basis accident (DBA) conditions. In addition; the system is subjected to DBA pressure (43 psig) at tempera-tures,up to 120r (per Surveillance Procedure SP623.13,' Rev.3) during Type A tests.
o he RBCCW system is seismically designed (per UFSAR Table-3.2-1) to the criteria in effect when the plant was licensed.
% e closed loop inside containment is not overpressure o
protected. his should not be a problem (as far as loop integrity is concerned) due to the short duration of high temperatures during a DBA.
3.
Because the RBCCW system can be considered a closed loop, it'only-requires single containment isolation va16es. We RBCCW syster.has valves which can act as containment isolation valves. %ese velves -
are check valve RC-6 and MOV gate valve RC-15. In its resolution cf SEP Topic VI-4, the NRC stated that the check valve inside contain-e ment at X-23 was not gencrally acceptable as a containment config-i i
uration (with respect to 10CFR50, Appendix A, General Design Cri-teria 54-57). his request supersedes our previous commitment to i
add new containment valves inside and outside the drywell at X-23
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and X-24.
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%e RBCCW system has the capability of performing some safety related functions in addition to essential functions (e.g., re-i circulation pump bearing and seal cooling is an essential function).
3 Its safety-related functions are containment heat removal (via c.on-1 tainment air recirculation and cooling units) and core decay heat a
i removal (via shutdown cooling heat exchangers which are cooled by RBCCW; shutdown cooling pumps are also cooled by RBCCW). Because of 1
these safety-related functions, the RBCCW system can be operating.
during all or part of a containment isolation incident (probably not 1
until last stages of recovery from a DBA because of potential over-loading of air recirculation fan motors in a 43 psig environment, loss ~of recirculation lines, and low reactor vessel level in early i
stages of a DBA). MECO believes that the existing containment valve configurations at X-23 and X-24 (inside check valve, and i
i outside remotely manually actuated gate valve) are consistent wi.th these system operational objectives.
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512132/Page 11 5.
When the RBCCW system pumps are operating and with valve 1-RC-15 i
normally open requiring reacte actuation for closure, the RBCCW
. system within containment will be at a pressure greater than.
containment atmospheric pressure during a dea (this is true even'if the loop inside containment is not functional; i.e., 1-RC-15 shut with RBCCW pumps running)s During a loss of normal power (WP) incident, emergency power is automatically aligned to the RBCCW pumps. h e pumps will be tripped by the WP and require manual restart by operator action.in the al 4
control room. 'Per procedure ONP 503B (on recovery from WP), opera-l tors are required to start one RBCCW pump immediately after an WP-incident. Werefore, the RBCCW system will be pressurized (to'a j
pressure >Pa) for all but.10 minutes of a simultaneous DBA and WP.
L ror these reasons, teECO concludes there are accident scenarios in which the RBCCW closed loop will be at a greater pressure than 1
containment DBA pressure. Leakage out of containment through X-23
.i and X-24 is not possible in such circumstances.
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If plant operators do not consider it prudent to operate the RBCCW system during all DaA recovery, operations, leakage out of contain-ment is not considered a threat to the health and safety oC the i
public for the following reasons:
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i Leakage of containment atmosphere into the RBCCW system o
(assuming-failure in the inside closed loop) will be diluted.
It mist travel through the closed system outside containment to i
the vented surge tank before it can be released.
It is very i
likely that gas leakage will be trapped in high points of the system before it can reach the surge tank.
CCNCWSION i
he items listed above indicate that it is not necessary to rely on containment valves in X-23 and X-24 to limit offsite dosage during a DBA.
i If the subje:t valves are not relied upon to accomplish containment is-2 olation or mitigate the offsite consequences of a DBA, they should not be subject to Appendix J testing. terefore, modifications to accomplish Appendix J testing should not be required at X-23 and X-24.
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B12132/Page 12 9
PLANT:
Millstone Unit No. 1 TITLE:
Exemption from 10CFR50, Appendix J, in Penetration X-15 DESCRIPTION NNECO requests exemption from Type C testing of check valve CU-29 in this reactor water cleanup (RWCU) system penetration.
BACKGROUND Penetration X-15 complies with Appendix A, General Design Criteria 54-57, as far as penetration valve configuration is concerned. However, only the outside valve (motor-operated gate valve CU-28) is Appendix J Type C testable with the existing piping configuration.
his exemption request is based on the premise that CU-28 is a t..sted, acceptable containment boundary.
JUSTIFICATION BASIS 1.
Check valve CU-29 is inside the drywell. Modifications to permit Type C testing (not even considering the testing itself) would l
result in 17.0 man-rem of personnel exposure. Wese modifications would cost approximately $482,000. A probabilistic risk assessment of this penetration has disclosed there is essentially no chance of any measurable public exposure due to leakage through this pene-tration.
2.
The RNCU system is seismically designed from the penetration to the cleanup recirculation pumps. This section of the system demon-strates its leak tightness continuously during plant operation at reactor coolant system temperature and pressure. Wis portion of the system also has two more check valves in each branch off of l
penetration X-15.
%ese valves will help to prevent leakage out of i
containment if CU-28 and CU-29 leak.
j 1
CONCLUSIONS NNECO requests exemption from Type C testing of check valve CU-29 on these bases.
2 I
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B12132/Page 13 PLANT:
Millstone Unit No. 1 TITLE themption from 10CFR50, Appendix J in Torus Water Sealed Penetrations X-204, A, B, and C, Ltnd X-210 A and B DESCRIPTION NNECO requests exemption from Type C testing in torus water sealed pene-trations. These penetrations are emergency core cooling system (ECCS) tie-ins to the suppression chamber (torus).
BACKGROUND All of the subject penetrations are part of closed loop systems. As such, only single containment isolation valves are required at each penetration (per 10CTR50,' Appendix A, General Design Criterion 57). This single valve design was acknowledged end accepted by the NRC in letter LS05-85-05-012 from J. A. Zwolinski to J. F. Opeka, dated May 10, 1985 (referen:e Technical Evaluation Report on "Containment TAakage Rate Testing", Section 3.1.3.4).
he exemption request supersedes a previous commitment to make modifica-tions to permit water seal testing in these penetrations.
JUSTIFICATION BASES 1.
A probabilistic risk assessment of the subject penetrations has disclosed that the conservatively calculated public safety benefit to be gained by leak testing these valves would be in the range of 0.42 man rem (this is assuming a leak in any one of these valves).
An estimate has disclosed it will take 17.0 man-rem and $3,051,000 to make modifications in these penetrations to accomplish Type C testing. his does not include the man-rem associated with performance of the testing and the cost of a water pressurization test rig (approximately $162,000). his expenditure of occupational exposure outweighs any small reduction in probablistic public man-rem to be gained.
2.
Penetrations X-204 A, B, and C, and X-210 A and B are for the low pressure coolant injection / containment cooling (LPCI) and core spray (CS) systems.
Piping through these penetrations remains open, or opens to containment after a design basis accident (DBA) or a main steam line break. Therefore, these valves do not act as containment barriers during the early stages of these accidents. %ese lines are left open to the containment during integrated leak rate tests and are therefore subjected to Type A leak testing at 43 psig.
3.
For a closed water sealed containment valve to act as a corus water depletion path, a leak must develop in the LPCI and CS piping out-side of the containment barriers.
4 m
m B12132/Page 14 The leak must be in the range of 11.3 to 12.9 gallons per minute to threaten drainage of the watu seals above the subject containment valves. We condition of LPCI and CS system packings and seals is visually checked each month during system functional tests.
It is unlikely that a packing or seal leak of the magnitude necessaty to deplete water seals in 30 days can occur without warning between these monthly tests.
In addition to the periodic testing these closed loopt. get, it should be pointed out that these lines are continuously exposed to a pressurized volume of water.
It can be shown that piping in lower sections of the CS & LPCI systems from the containment valves back to the penetrations (X-204 A, B, and C only), is re.taining water at 4.8 psig to 6.5 psig.
(Torus minimum water level is at El. -10';
LP-2A and 2D are at El. -25'; LP-2B and 2C are,at El. -22'1"; CS-2A.
and 2B are at El. -22'30"; LP-24A through D are at El. -21').
We "keep fill system" keeps the upper sections of the CS and LPCI systems filled and pressurized to 150 psig and 45 psig respectively.
Wese water seals demanstrate the leak tightness of gaskets and packings on a continuous basis.
4.
With pressure on their downstream sides, the containment valves in question can not experience the differential pressure imposed during-an Appendix J Type C water seal test (i.e., 47.3 psid). Consequent-ly, these water sealed containment valves can never leak as m2ch in a real post-accident situation as would be measured to comply with Appendix J.
5.
Valve packing and pump seal leakage (passive failures), as discussed above, would be a function of the post-accident containment tempera-ture and pressure, and head losses between the water sealed valves and the presumed discharge point. None of these parameters have any relationship to the water tv.ts imposed by 10CFR50, Appendi:( J l
(e.g., AP and temperature of test differ from post-accident condi-tions; Appendix J tests would be done with minimal downstream line loss).
j 6.
Per FSAR Figure 5.2.3.3.4, post-accident pressures drop to approximately normal operating pressures (i.e., 1.75 psig) in 11.6 days.
Setting a water leakage limit for torus containment isolation valves, based on depleting the torus water seal over these valves after 30 days of exposure to 47.3 psid, is not representative of post-accident conditions.
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B12132/Page 15 7.
To perform Appendix J water seal tests, it would be necessary to shut the subject containment valves and adjacent valves to set up test pressurization boundaries. Many of the non-containment valves to be closed as test boundary valves are large, rarely used, un-tested, butterfly valves likely to have Jarge leakage rates.
It is possible that test boundary valve leakage may be large enough to throw off or prevent performance of a containment valve's test.
Wis would require repairs that could only be accomplished by drain-ing the torus. his type of maintenance would be highly undesirable because it would not be related to the containment valves themselves (i.e., this maintenance would have no bearing on the success of the subsequent containwnt valve test).
8.
Some of the modifications required to accomplish Appendix J water seal testing will require rerouting of plant piping. his could invalidate much of the torus attached piping analysis that has been cenpleted to date.
%ere is also a problem in finding the room to add all of the new piping and valves (test pressure boundary valves, vents and drains, possibly expansion loops, etc.) that would be required to accomplish water seal testing.
9.
Were are means of replenishing the torus water level before it is depleted (e.g., frca the condensate storage tank, sea water, keep fill system back leakage). Not all of these means involve Category I seismically designed systems that are accessible after a DBA or use of demineralized water. However, the possibility of replenishing the torus water should be acknowledged in consideration of this exenption request.
CCNCWSION In conclusion, the torus is considered an uMepletable water source.
Modifications to demonstrate that a 30 day water seal can be maintained have no real safety benefits.
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