ML20024G370
| ML20024G370 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/30/1976 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G368 | List: |
| References | |
| NUDOCS 9102110384 | |
| Download: ML20024G370 (34) | |
Text
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RLVZSION W.
llegub}0gDOCkBtfjl EXHIBIT A MONTICELLO NUCLEAR GENERATING PIANT DOCKET NO. 50-263 LIC MSE AMENDMENT-REQUEST DATED JANUARY 30, 1976 PROPOSED CHANCES TO THE TECHNICAL SPECIFICATIONS, APPENDIX A 0F PROVISIONAL OPERATING LICESE DPR-22 Pursuant to 10 CFR 50.59, the holders of Provisional Operating License
'DPR-22 hereby propose the following changes to the Appendix A Technical Specific ations :-
Specification and Bases 3.7/4.7. A - Primary Containment PROPOSED CHANGES
-1.
Add new Limiting Conditions for Operation and Surveillance Require-ments for primary containment leakage testing.
We proposed new specifications appear on pages 140 through 143 of Exhibit B.
i 2.
Redesignate existing specifications 3/4.7. A.3 through 3/4. 7. A.6 as specifications 3.7/4.7. A.5 through 3. 7/4. 7. A. 8 Wese changes appear on pages 144 through 148 of Exhibit B.
Page 147A will be deleted by these changes.
3.
Revise Table 3.7.1 to include automatic isolation valves smaller than 2 inches in diameter which were omitted from the original-table. he proposed new table appears on pages 153,154, and 154A of Exhibit B.
'4.
Add new Table 4.7.1, "Monticello Containment Penetrations," as shown on pages 154B through 154 O of Exhibit B.
n is table lists all Type B
- and Type C testing requirements.
5.
Revise. the " List of Tables" in the front of the Monticello Technical Specifications to reflect the change in the title of Table -3.7.1 and the' inclusion of the new Table 4.7.1.
REASONS FOR CHANCES In a letter from Mr. K. R. Go11er, Division of Reactor Licensing, USNRC, to Mr. L. O. Mayer, NSP, dated August.13, 1975, NSP was requested to determine if containment leakage testing at the Monticello Nuclear Generating Plant conforms to 10 CFR 50 Appendix J..
NSP was specifically asked to_ identify any _ design = features that do not permit conformance with the requirements of l
Appendix J or any existing Technical Specifications that are less restrictive than Appendix J. - A preliminary response to this request was contained in a letter from Mr. L. O. Mayer, NSP, to Mr. K. R. Coller, USNRC, dated g
September 19,- 1975.- n is letter _ outlined the following actions and schedule 1
9102110384 760504 PDR-ADOCK 05000263 P
PDR i.
_=-
REVISION MAY k, 1976 EXHIBIT A 2
to attain conformance to Appendix J:
a.
A License Amendment Request to be submitted by December 31, 1975 (later rescheduled for January 30, 1976) to revise the Monticello Technical Specifications to conform to Appendix J in those areas where plant design permits.
b.
An analysis, to be submitted by March 31. 1976, of systems contain-ing isolation valves which require Type C tests in accordance with the definition contained in Section II.H of Appendix J, but which are not testable. Appropriate design changes will be proposed or a request for exemption from the requirements of Appendix J will be included in conformance with 10 CPR 50.12.
The Technical Specification changes proposed in this License Amendment Request satisfy she first of the tuo commitments made in our letter of September 19, 1975.
The proposed changes contained in Exhibit B revise the Technical Specifi-cations to remove conflicts with 10 CFR 50, Appendix J.
The proposed wording is similar to the wording used in the Technical Specifications of recently licensed BWR's conforming to Append x J.
With the exception of the following clarifying remarks, no further justification for these changes is necessary, Proposed specification 4.7. A.3.a permits NSP to determine a.
an Le value for the Monticello containment at the next refueling outage mad to conduct future leakage tests at a reduced pressure of Pe=0.5Pa.
This Le determination is normally completed during the initial integrated leakage test conducted as part of a plant's preoperational testing program.
Existing Technical Specifications have required testing at Pa; therefore an Lt value was not determined during the initial test at Monticello.
For purposes of scheduling future integrated leakage tests, the test con-ducted during the next refueling outage would constitute the preoperational Type A leakage rate test specified in Appendix J.
b.
Proposed specifications 3. 7. A.3.C and 4. 7. A.3. f would permit testing of main steam isolation valves. at 25 psig. As dis-cussed in.the proposed 4.7 Bases on page 163 of Exhibit B, testing of these valves at Pa is not feasible and there is no substantial benefit to be gained from testing at Pa*
Reduced pressure testing of these valves is considered a departure from Appendix J.
A request for exemption from this requirement will be submitted with the report due March 31,-1976.
b REVISION 4
EXHIBIT A MAY 4, 1976 3
Proposed specification 4.7. A.4.a permits reduced pressure overall c.
testing of the air lock every three days when the air lock is in use.
This is a reasonable surveillance requirement to verify correct door sealing when the air lock is actually in use. Casket leakage tests are not possible since Monticello air lock doors are not equipped with double gaskets.
i Reduced pressure testing of the air lock is considered a departure from Appendix J.
A request for exemption from this mquirement will be submitted with the report due March 31, 1976 d.
Table 3.7.1 has been revised to include automatic isolation valves in containment penetrations smaller than two inches in diameter.
These penetrations were omitted from the original table.
Table 3.7.1 has also been revised to list isolation valve identification numbers, e.
Table 4.7.1 is a new table that has been included in the preposed Technical Specification changes as a guide in perform-ing Type B and Type C tests. All containment penetrations are listed along with the sealing device or isolation valves in each penetration and the testing required for endi.
In cases where conflicts in the table exist and a Type C test is specified for a valve which is not testable in place, the valve is identified as not testable.
Each conflict will be resolved by approval of a request for exemption from a requirement of Appendix J or completion of a plant modification to permit testing.
A request for exemption from certain requirements of Appendix J and a description of planned modifications will be submitted to the Commission for review and approval.
SAFETY EVALUATION This License Amendment Request is being submitted at the request of the Commission to remove -the current conflicts between the surveillance re-
-quirements for containment leakage testing in the Monticello Technical
'--- Specifications and in 10 CFR 50,- Appendix 1. - The proposed changes re-vise the Monticello Technical. Specifications to conform to Appendix J in all areas where plant design permits.
Where possible, the proposed changes follow the wording used in Technical Specifications issued for BWR's currently being licensed and whose testing programs conform to Appendix J.
j a
REVISION MAY 4, 1976 LICENSE AMENIME!Tt REQUEST LATED JANUARY 30, 1976 EXHIBIT B Ibis exhibit consists of the following pages revised or added to incorporate all of the proposed Technical Specification changes:
vii 140 through 148 l
153 154 154 A through 154 0 (newpages) 162 through 164 Existing page 147 A is to be deleted by these changes.
l I
1 i
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'<h A
3 LIST OF TABLES 3.1.1 Reactor Protection System (Scram) Instrument Requirements 30 d
k 4.1.1 Scram Instrument Functional Tests - Minimum Functional T-st Frequencies 34 i
for Safety Instrumentation nnd Control Circuits Ti
^
4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for 36 B
Reactor Protection Instrument Channels
+
h 3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 50 i iD 3.2.2 Instrumentation that Initiates Emergency Core Cooling 53 r
Systems 3
,j 3.2.3 Instrumentation that Initiates Rod Block 57 w
3.2.4 Instrumentation that Initiates Reactor Building Ventilation 60 l
Isolation and Standby Gas Treatment System Initiation l
3.2.S Trip Functions and Deviations 69 4.2.1 Minia n Test and Calibration Frequency For Core Cooling, Rod Block 61 and Isolation Instrumentation 3.6.1 Safety Related Ilydraulic Snubbers 121B i
4.6.1 In-Service Inspection Requirements for Monticello 124 g
w i
3.7.1 Primary Containment Automatic Isolation Valves 153 fr 4.7.1 Monticello Containment Penetrations 154B m
h.8.1 Sample Collection and Analysis Monticello Nuclear Plant - H1vircr= ental 174 Monitoring Program 6.5 1 Protection Factors for Respirators 206 vii PEV O $i
m 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIATE REQUIREMDfTS (d) During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal cooldown rates if the torus water temperature exceeds 120*F.
(e) MinL n Water Volume 68,000 cubic 1
feet.
(f) Maximum Water Volume 77,970 cubic feet.
2.
Prior to establishing conditions requirin6 primary 2.
Primary containment integrity as defined in containment integrity verify:
the Section 1, shall be maintained at all times when the reactor is critical or when the reactor water temperature is above
- a. All penetrations not capable of bein6 closed 212*F and fuel is in the reactor vessel ex-by operable containment automatic isolation cept while perfoming low power physica valves and required to be closed during
. tests at atmospheric pressure during or accident conditions are closed by valves, after reibeling at power levels not to ex-blind flanges, or deactivated automatic i
l ceet 5 Nw(t).
"alves secured in position.
3 Concainment leakn6e rates shall be limited to:
- b. All equipment hatches are closed.
- a. An ovem11 integrated leakage rate of:
- c. Both containment air lock doors are closed.
- 1. 5 La (1.2 percent by weight of the 3
The containment leakage rates shall be denunstrated contain.nent air per 2h hours) at Pa at the following test schedule and shall be determined (k1 psig), or in confomance with the criteria specified in Appendix J of 10CFE50 using the methods recorOwnded
- 5 in ANSI IAS.h - 1972
4
- 2. $ Lt at a Irduced pressure of Pt (20 5 ps!g).
f
- b. A combined leakage rate of 0.6In for all k*
penetrations and valves (except main steam isolation valves) subject to Type B and-C tests when pressurized to Pa-140 37/h.7
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3.0 LIMITING CONDITIONS mR OFERATION 4.0 SURVEILLANCE RDQUIFIMEITfS
- a. During the first refueling outage following c.115 scf per hour for any one main steam isolation the adoption of this specification, a Type A -
valve when tested at 25 psig.
test shall be perfomed at a pressure Pt of 20 5 psig and a secend test performed at a pressure Pa of 41 psig. 'Ibe maximum With either (a) the overall integmted containment allowable test leanage rate, L, shall be leakage mte exceeding 0.75Ia or O.75L, as t
t deterained in accortlance with section applicable, or (b) with the measumd combined III.A..k(a)(iii) of Appendix J.
leakage rate for all penetrations and valves subject to Type B and C tests exceed *. g 0.6L,
or (c) a main steam isolation valve leak rate
- b. Following the test specified in 4.7.A.3.s-a above, three Type A tests shal1 be conducted exceeding 11.5 scf per hour, restore the leakage at 40 + 10 month intervals during shutdown rate (s) to within acceptable limit (s) prior to at eitiier Pa or P during each 10-year t
increasing the reactor coolant tempemture above 2120F.
service period. One of these tests shall be conducted during the shutdown for each 10-year plant inservice inspection.
- c. If any Type A test fails to meet the acceptance criteria of 0 75Lt for reduced pressure tests at P r O.7% for peak pmssum t
tests at P, the test schedule for sub-a sequent Type A tests shall be reviewed and approved by the Commission. If two such consecutive tests fall to meet the acceptance criteria, a Type A test shall f
be perfomed at least every 18 months until two consecutive Type A tests meet g
the acceptance criteria at which time g
the schedule specified in 4.7.A.3.b y
may be resumed.
l 3 7/4 7 161 REv l
3.0 LIMITIliG CONDITIONS FOR OPERATION k.0 SURVEILIANCE RDQUIPIMEICS
- d. 'lhe accuracy of each Type A test shall be verified by a supplemental test which:
- 1. Confirms the accuracy of the test by verifying that the difference between the supplemental data and the Type A test data is within O.25Lt for reduced pressure tests at Pt or within 0.25I,
for peak pressure tests at P -
a
- 2. Is of sufficient duration to establish accurately the change in leakage rate between the Type A test and the supplemental test.
- 3. Requires the rate that 6as is in) eted into the contair: ment or bled from the containment during the supplemental test to be equivalent to at least 25 percent of the leakage rate mea-sured during the Type A test.
- e. Type B and Type C tests shall be conducted f1 at intervals no greater than 24 months except for tests of the containment air f
lock, and shall include all testable w
co=ponents listed in Table 4.7 1.
- f. Type B and Type C tests shall be conducted at Pa, except for main steam isolation valves and the air lock.
i 3 7/4.7
- 8 i
arv 1
30 LIMITING CONDITION 3 FOR OPERATION h.O SURVEILLANCE REQUIRDMS
- 4. The containment air lock eball be demonstrated k.
The containment air lock shall be operable with:
operable by:
a.
Both doors closed except when the air lock is being used for norml entry and exit, then at
- a. At least once per 6 months by conducting an least one air lock door shall be closed.
overall air lock leakage test at 10 psig and by verifying that the overall leakage rate is
- b. An overall air lock leakage rate of L O.OSIo at within its limit.
If the air lock is in use, at 10 psig.
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and containment integrity is required, the air lock shall be tested every three days our
- c. All interlocks function as designed.
after each use, whichever interval is greater.
- b. During each refueling outage and following repairt to the air lock, by verifying all interlocks function as designed.
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IA lh3 3 7/h.7 aEv
30 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMEIffS 5
Pressure Suppression Chamber -
- 5. Pressure Suppression Chamber -
Reactor Building vacuum Breakers Reactor Building Vacuum Breakers a.
Except as specified in 3.7. A. 5.b a.
The pressure suppression chamber-reactor below, two pressttre suppression building vacuum breakers and associated in-chamber-reactor building vacuum strumentation including set point shal' be breakers shall be operable at all checked for proper operation every three times when the primary containment months.
integrity is required. The set point of the differential pressure instrumentation which actuates the pressure suppression chamber-reactor building vacuum breakers shall be 0.5 pai.
b.
From and after the date that one of the pressure suppression chamber-reactor building vacuum breakera is made or foun' to be inoperable for any reason, reactor operation is permissible only during the suceed-ing seven days unless such vacuum breaker is soonez made operable, g
provided that the repair procedure x
does not violate primary containment e
integrity.
M 3.7/4.7 ing REv
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SIIRVEILLANCE REQUIREMENTS 6.
Pressure Suppression Chamber-Drywell Vacuum 6.
Pressure Suppression Chamber-Drywell Breakers Vacuum Brea'.ers I
a.
When primary containment is required, all a.
Operability and full closure of the drywell-suppression chamber vacuum breakers drywell-suppression chamber vacuum shall be operable and positioned in the breakers shall be verified by performance closed position as indicated by the of the following:
position indication pystem, except during testing and except as specified in 3.7.A.
(1) Monthly each operable drywell-6.b and c below.
suppression chamber vacuum breaker shall be exercised through b.
Any drywell-suppression chamber vacuum an opening-closing cycle.
breaker may be nonfully closed as indicated by the position indication and (2) Once each operating fuei' cycle,.
alarm systems provided that drywell to dryvell to suppression chamber leakage suppression chamber differential pressure shall be demonstrated to be less decay does not exceed that shown on Figure than that equivalent to a one-inch 3.7.1.
diameter orifice and each vacuum breaker shall be visually inspected.
c.
Up to two drywell-suppreursoa chamber (Containment access required)
+
vacuum br*Aars may be inoperable provided that: (1) the vacuum breakers (3) Once each operating cycle, vacuum are determined to be fully closed and at breaker position indication and least one position alarm circuit is alarm systems shall be calibrated and operable or (2) the vacuum breaker is functionally tested.
(Containment secured in the closed position.
access required)
- f x
(4) Once each operating cycle, the y
vacuum brukers shall be tested to determina that the force required to k open each valve from fully closed to o
fully open does not exceed that equivalent to 0.5 psi acting on the suppression chamber face of the valve disc. (Containment access required) 145 3.7/4.7 REV
%D SURVEILLANCE REQUIREMENTS 3.0 LIMITING CONDITIONS FOR OPERATION b.
When the position of any dryvell-d.
One p eltion alarm circuit can be in3perable providing that the redundant position alarm suppressi?n chamber vacuum breaker valve circuit is operable. Both position alerm is indicated to be not-fully closed at a time when cuch closure is required, the circuits may be inoperable for a period not dryvell to suppression chamber differential to exceed seven days provided that all vacuum pressure decay shall be demonscrated breakers are operable.
to be less chac that shown on Figure 3.7.1 insediately and following any evidence of eubsequent operation of the inoperable valve until the icoperable valve is restored to s normal condition.
When both position alarm circuits are made c.
or found to be icoperable, the control panel indicator light status shall be recorded daily to detect changes in the vacuum breaket position.
I 7
Oxygen Concentration 7
Oxygen Concentration Whenever inerting is required, the primary After completion of the startup test a.
containment oxygen concentration shall be program and demonstration of plant measured and recorded on a weekly basis.
electrical output, the primary contain-i l
ment atmosphere shall be reduced to less than 51, oxygen with nitrogen gas whenever the reactor is in the run g:
moda, except as specified in 3.7. A.7.b.
M
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b.
Within the 24-hour period subsequent to placing the reactor in the run mode p
following shutdown, the containment atmosphere oxygen concentration shall be I
redrced to less than 5% by weight, and maintained in this condition. Deinerting may conenence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to leaving the run mode for a
reactor shutdo m.
146 3 7/4.7 PB
+
e 30 LIMITING CONDITIOfE FOR OPEPATION 4.0 st-.lILIANCE REQUIREMENTS
- 8. If any of the specifications of 3 7.A cannot be met, the reactor shall be placed in the cold shutdown condition 'within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 b
if w
F 0E 3 7/h.7 147 RE7
9 I
J 30 LIMITING CONDITIONS WR OPERATION h.0 SURVEII.IANCE REEUIREMENTS j
B.
Standby Gas Treatment System B.
Standby Cas Treatment Systeta
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1.
Dccept as specified in 3 7.B.3 below,
"" I 8**
- I*"***
both circuits of the standby gas t: eat-
- P*#
"I ment system shall be operable at all
.)
times when secondary containment g
g, gg integrity is required.
shall be demonstrated that:
(1) Pressure drop across the combined high-efficiency and charcoal filters is less than 7.0 inches of water, and (2) Inlet heater output is at least 15 kw.
- b. Within 30 days of the beginning of each refueling outage, whenever a filter is changed whenever work is performed that could affect filter systems efficiency, and ac l
intervals not to exceed six months between l
refueling outages, it shall be demonstrated that:
j (1) The removal efficiency of the installed particulate filters for particles having a mean diameter of 0.7 microna shall be k
x
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148 g
E 37/h.7
Table 3 7.1 Primary Containment Automatic Isolation Valves Pemissible Isolation Group Isolation Valves Operating Nomal (Note 1)
Application Inboani Outboard Time (sec)
Pccition 1
Main Steam Isolation AO-2-80A A0-2-86A 3 <t< 5 Open A0-2-80B AO-2-86B
~~
AO-2-80C AO-2-86C A0-2-80D AO-2-86D lhin Steam Line Drain MO-2373 MO-2374 t 160 Closed Reactor Water Sample CV-2790 CV-2791 t <60 Closed 2
Drywell Fquipment Sump AU-2561A A0-2561B t $60 Open Drywell Floor Sump AO-25hlA A0-2541B t 160 Open CV-2384 t <60 Open(Ibte 3)
Torus Vent Bypass Torus to N2 Recirculation CV-7440 t 560 Open(Ibte 3)
Torus Vent A0-2383 t 160 Closed Torus Vent AO-2896 t <60 Closed Drywell Vent Dypass CV-2385 t $60 Closed Drywell Vent AO-2386 t 160 Closed Drywell Vent AO-2387 t $60 Closed
'Ibnts Air Purge Air Supply A0-2378 t 160 Closed Drywell Air Purge Supply Ao-2381 t 160 Open(Note 3)
E Containment Air Purge Supply AO-2377 t $60 Open(Ibte 3) f-TIP Ball Valves (3)
!bte 2 Closed eN Nitrogen Pumpback Suction CV-7h36 t $60 Open Nitrogen Pumpback Suction CV-7h37 t $60 Open 153 3 7/h.7 Fr/
Table 3 7 1 Primary Containment-Automatic Isolation Valves (continued) f Pennlasible Isolation Group Isolation Valves Ope nting Nomal (Note 1)
Application Inbonni Outboard Time'(sec)
Position 2
RHR Supply MO-2029 Mo-2030 t 1120 Closed RHR Head Cooling MO-2027 MO-2026 t $120 Closed RHR Return to A Loop MO-2014 t $120 Closed RHR Return to B Loop MO-2015 t $120 Closed Containment Nitmgen Supply CV-3269 t 160 Closed Torus Nitrogen Supply CV-3267 t 160 Closed Drywell Nitrogen Supply CV-3268 t10 Closed 6
Oxygen Analyzer Sample Point CV-3305 t 160 open Oxygen Analyzer Sample Point CV-3306 t 160 Open Oxygen Analyzer Sample Point CV-3307 t 160 open Oxygen Analyzer Sample Point CV-3308 t 160 open Oxygen Analyzer Sample Point CV-3309 t 160 open 4
]
Oxygen Analyzer Sample Point CV-3310 t$0 open 6
Y Oxygen Analyzer Sample Point CV-3311 t <60 Open x
Oxygen Analyzer Sample Point CV-3312 t $60 Open m
Oxygen Analyzer Return CV-3313 t10 Open M
6 i
j 0xygen Analyzer Return CV-3314 t 16o Open i
j 37/h.7 154 azV
Table 3 7.1-EPrimary Containment Automatic Isolation Valves (continued)
Isolation Group-Pemissible (Note 1)
'IsolationiValves'
. Operating '
Nomal
. Application.
Inboani
' Outboard Time (sec)
, Position 3
Reactor water Cleanup Supply MO-2397 MO-2398 t sho'
'open-Beactor Water Cleanup Return
'M0-2399 t 140 Open 4
HPCI Steam Supply MO-2034 MO-2035 t $ho open 5
RCIC. Steam Supply MO-2075'
.MO-2076
.t $30 open Note 1:
Containment isolation gmupings are as follows:
Group 1 The valves in Group 'l are closed upon any one of the following conditions:
- 1. Reactor lov low water level
- 2. Main steam line high radiation
- 3. Main steam line high flow
- 4. Main steam line tunnel high' temperature
- 5. Miin steam line' low pressure (RUN mode only)
Group 2 The valves in Gmup 2 are closed upon any on-of the following conditions:
- 1. Reactor low ' vater level
- 2. High Drywell Pressure Group 3 The actions in Group 3 are initiated by reactor lov vater level.
Group h Isolation valves in the -HPCI System are closed upon any one of the following conditions:
- 1. HPCI steam line high flow
- 3. High temperature in the vicinity of the HPCI steam line.
- 2. HPCI steam line low pressure
.f Group 5 Isolation' valves in the RCIC System are closed upon any oae of the following conditions:
~
- 1. RCIC steam line high flov
. 3. High temperature in the vicinity of the RCIC steam line.
-E.
- 2. RCIC steam line low pressure Y
Eote. 2:
Testing consists of verifying TIP automatic withdrawal and. ball valve closum on a simulated Group 2 isolation signal.
Note 3:
'lhese valves are open when the nitrogen recirculation system is in service and during nitrogen addition.
37/4.7 154A REV
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1
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Table k.7 1 tbnticello Containment Penetrations Applicable Appendix IIMER BARRIER Ourst< BARRIER Dasignation Description J Type Test Designation Type Testable Designation Type Testable Penetration Seismic Restraint B
1 Yes Port A
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l Ye8
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Seismic Restraint B
Port B Seismic Restraint B
1 Yes
, Fort C Seismic Restraint B
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1 Y*8 Port D 1
Yes Seismic Restraint B
Port E 1
Yes Seismic Rectraint B
Port F Seismic Restruint B
1 Yes Port G 1
Yes Seismic Restraint B
Port H
't Dryvell Head B
1 Yes F
55 15hr 3 7/4.7 REV
o Table h.7.1 M;nticello Containment Penetrations (continued)
INNER BARR 3 OUTER BARRIER Applicable Appendix Penetration Desigration Det. cription J Test Type Designation Type Testable Designation Wpe Testable 1
Yes X-1 Equipment Hatch B
Yes X-2 Air Inck (Note 6) 3
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i X-3 Not Assigned NONE Xh Head Access Fatch B
1 Yes X-5A - 5H Dove n -Torus NONE (Note 7)
Vent Pipes X-6 CRD Access Hatch B
1 Yes X-7A Bellows B
2 Yes Primary Steam Line A C (Note 1)
AO-2-80A 3
Yes AO-2-86A 3
Yes X-7B Bellows i
B 2
Yes Primary Steam Line B C (Note 1)
A0-2-803 3
Yes Ao-2-86B 3
Ye8 X-7C Benous B
2 Yes M
1 Primary Steam Line C C (Note 1)
AO-2-800 3
Yes AO-2-86c 3
Yes 1
y 2
Yes y
X-7D Be novs B
Primary Steam Line D C (Note 1)
AO-2-80D 3
Yes A0-2-86D 3
Yes 2
Yes X-8 Ben ovs B
I Primary Steam Drain C
MO-2373 4
Yes MO-2374 4
Yes 37/h.7 154c P5 wwn u
Table 4 7 1 lbnticello Containment Penetrations (continued)
A plicable Appendix INNER BARRIER OtTIER BARRIER P
Penetration Designation Description J Type Test Designation Type Testable Designation D
Testable X-9A Bellows B
2 Yes Feedvater Line C
W-97-2 5
Yes W-94-2 5
Yes
]
2 Yes X-9B Bellows B
l W-97-1 5
Yes W-9k-1 5
Yes Feedvater Line C
2 Yes X-10 Bellows B
MO-2075 h
Yes MO-2076 h
Yes Steam to RCIC C
X-11 Bellows B
2 Yes Steam to HPCI MO-20) 4 Yes MO-2035 4
Yes C
X-12 Bellows 2
Yes B
RHR Supply MO-2029 4
Yes MO-2030 4
Yes C
2 Yes X-13A Bellows B
9 s
' ~ ~ '
LPCI to B loop C
MO-2015 4
Yes MO-2013 h
Yes 5
X-13B Bellows B
2 Yes
?
LPCI to A loop C
MO-2014 h
Yes MO-2012 h
Yes ve I
~s I
Bellows-B j
X.14 2
Yes MD ?WWD C
MO-2397 4
Yes MO-2398 4
Yes yn 4 X-15 TM^;w;-,w.
%e.1%instrafion f
as -
ya 17 ra m
- a+
-; }:y
- p repa+; e3 w.
i! l3 l_%>
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- seeyge,
l I
l Table 4.7.1 Fbnticello containment Penetrations (continued) l r
l Penetration Applicable Appendix INNER BARRIER OUTER BARRIER Designatio't Description J Type Test Designation Type Pestable I;esignation Wpe Destable l
X-16A Bellows B
2 Yes Core Spray B C
MO-175h 4
Yea MO-1752.
h Yes X-16B Bellows B
2 Yes Core Spray A C
MO-1753 h
Yes MO-1751 h
Ye8 X-17 Bellows B
2 Yes Head Cooling C
MO-2027 4
Yes MO-2026 4
Yes X-18 Floor Sump Discharge C
AO-25hlA 7
Yes i
A0-2541B 7
Yes X-19 Equip Sump Discharge C
AO-2561A 7
Yes A0-2561B 7
Yes X-20 Demin Water Supply c
m-57 8-no m-58 8
no X-21 Service Air Supply C
AS-39 8
no AS-hO 8
no y
M X-22 Instrument Air C
CV-lh78 9
no t-X-23 RBCCW to Dryvell C
RBCC-15 5
No UN X-2h RBCCW from Dryvell C
MO-lk26 h
No 37/h.7 154E REV
Table h.71 Monticello Containment Penetrations (continued)
~
Penetration Applicable Appendix INNER BARRIER OLTTER BARRIER Desigration Description J Type Teat Designation Type Testable Designation Type Testable X-25 Dryvell Ventilation C
AO-2386 10 Yes Exhaust AO-2387 10 Yes cv-2385 9'
Yes X-26 Drywell Ventilation C
AO-2377 10 Yes Supply AO-2 10 Yes AO-2 1 10 Yes CV-3267 9
Yes CV-3263 9
Yes cV-3269 9
Yes X-27A - 27C Instrumentation NONE (Note 3) 18 X-27D Oxygen Analyzer C
CV-3305 9'
Yes Sample Pbint cV-3306 9
res X-27E Oxygen Analyzer C
CV-3307 9
Yes Sample Point Y
CV-3308 9
Yes l
F X-27F Oxygen Analyzer C
CV-3309 9
Yes s
Sample Point CV-3310 9
Yes i
3.7/h.7 154F REV
. 'lbble 4.7.1 Monticello Containment Penetrations (continued)
Penetration Applicable Appendix INNER BARRIER OUTER BARRIER Designation Description
'J Type Test Designation Type Testable Designation Type Testable.
X-28A - 28F Instrumentation NONE (Note 3) 18 X-29A - 29 D Instrumentation
' NONE (Note 3) 18 X-29E - 29F Instrumentation B
(Note 8) 19 No -
X-30A - 30F Spare Penetrations NONE 17 t
X-31A,B,D,E,F Instrumentation NONE (Note 3) 18 X-31C Spare NME 17-l X-32A,B,D,E,F Instrumentation NME (Note 3) 18 i
X-32C Dryvell Flood B
(Note 8) 19 No.
level Switch X-33A - 33F Instrumentation NONE (Note 3) 18 X-34A - 34F Spare Penetrations NONE 17 X-35A,B,C TIP Probes ' ( Note h )
TIP 1,2,3 15 No.
g.
C w.
X-35D Spare Penetration NONE 17 y
X-35E TIP Purge Supply
.C Purge Check 5
No os Purge SV 16 No' 3 7/h.7 154G REV
Table 4.71 tbnticello Containment Penetrations (continued)
Applicable Appendix INNER BARRIER OUTER BARRIER J Type Test Designation Type Testab17 Esignation Trpe
[bstable Penetration Designation Description C
CRD-3h 5
do CRD-31 5
Yes 4
Yes XR-26-2 5
Yes X-36 CRD Hydrsulic Return X-37A 12 Recire Seal Inj C
XR-27-2 X-37A - 37D CRD Insert Lines (121)
NNE (Note 5) c; XR-26-1
_5 Yes X-38A 11 Recire Seal Inj C
XR-27-1 Yes X-38A - 38D CRD Withdraw (121)
NONE (Note 5) 5 MO-2021 h
No C
X-39A Drywell Spray B MO-2023 4
No C
MO-2020 4
No X-39B Dryvell Spray A MO-2022 h
No 18 X-40AA - 40DF Instrumentation NONE (Note 3)
CV-2790 9
Yes CV-2791 9
Yes X-kl Recire loop B Sample C
XP-7 5
No XP-6 5
Yea X-42 Standby Liquid Contro]
C 17 X 47 Spare Penetrations NCNE
~
~
~
y-h8 Nitrogen Pumpback C
- )
Suction CV-7437 9
Yes 18 g
X 49A - 49F Instrwnentation NONE (Note 3) 18 B (Note 8) 19 No E
X-50A - 50D Instrumentation NNE Ofote 3)
X-50E - 50F Instrumentation 18 G
X-51A - 51F Instrumentation NWE (Note 3) 18 X-52A - 52F Instntmentation NONE (Note 3)
X X-99 Not Assigned NmE 15hH 37/h.7 Rsv CAL;
Table 4.7.1 Fbnticello Containment Penetrations (continued) s Penetration pplicable Appendix IriNER 1ARRIEF otrPRD nAporrn Designation Description J Type Test Designation Type
'Ibstable lesignation lype Testable X-100A - 100D Electrical Penetration B
11 Yes X-100E Spare Penetration NONE 17 17 X-101A,101C Spare Penetrations NONE h
11 Yes X-101B,101D Electrical Penetration B
X-102 Spare Penetration NONE 17 11 Yes X-103 Electrical Penetmtion B
X-104A-- 10hD Electrical Penetration B
11 Yes X-10hE Spare Penetration NONE 17 X-105A,105C, Electrical Penetration B
11 Ye8 lO5D X-lO5B Spa m Penetration NWB 17 X-106 Spa m Penetration NmE 17 X-lO7 Spare Penetration NONE 17 X-108 - X-199 Not Assigned NONE f
N X-2OOA Torus Hatch (h50)
B 1
Yes f
k X-2003 Torus Hatch (225 )
B 1
. Yes m
X-201A - 201H Ibrus Vent Pipes NONE (Note 7)
X-202A,B,C,D, Dryvell-1brus NGiE (Note 7)
E,F,G,B, Vacuum Breakers i
J,K l
15h I 3 7/h 7
- c. a. - a p.,
Table k.7.1 Manticello Containment Penetrations (continued) 1 Penetration Applicable Appendix INNER BARRIER OUTER BARRIER Designation Description J Type Test Designation Type Testable Designation Wpe Testable l
l X-202I Not Assigned NONE X_203 Not Assigned NQiE X-20bA - ?O4D Torus Ring Header NOiE (Note 7)
X-205 Torus Ventilation C
Ao-2383 lo Yes c)
Exhaust and Supply CV-2384 9
Yes I
to Nitrogen Becircu-AO-2696 10 Yes
- 1ation System CV-7440 9
Yes l
X-206A - 206D Ibrus Instrumentation B (Note 8) 19 g
X-207A - 207H Torus Vent Pipe Drains NONE (Note 7)
X-208A - 208H Belief Valve Discharge NONE (Note 7)
Pipes X-209A - 209D Torus Instrumentation B (Note 8) 19 No X-210A RHR and Core Spray B NOiE (Note 9)
RHR-8-2 5 ^
Test Line to Ibrus MO-2007 4
]
MO-2OO9 12 I
Mo-1750 12 M
f' CS-10-2 13 g
I-eN 3 7/4 7 1N REV r%.
b
.~.
^
Table 4.71 bnticello Containment Penetratidns (continued)!
OUTER BARRIER INNER BARRIER Applicable Appendix
! Penetstion -
J Type Test Designation Type Testable Designation' TiP6 Testable
! Designation Description s
[ X-210B RER t.nd Core Spray A NONE (Note 9)
PJIR-8-1 5
5 Test L1.w to '1brus MO-2006 4
5 3
t f.
MO-2OOB 12 t
MO-17h9 12 O
3 CS-10-1 13 1
~
1 MO-2OO7 4
No.
C X-211A RHR B Torus Spray MO-2OO9 12 fb i"
i MO-2011 12 tb MO-2006 4
No C
X-211B RHR A 'Ibrus Sprny MO-2OO9 12 1b No MO-2010 12 I'
RCIC-9 5
Yes X-212 RCIC Turbine Exhaust C
d 1
X-213A,213B Flanged Bottom Torus NONE (Note 10)
$w Drains CV-3313 9
Yes E
j X-21h Oxygen Analyzer Return C
g CV-3314 9
Yes a
4\\
17 5
f X-215 - 216 Spare Penetrations NWE NONE (Ibte 11)
E j X-217 HPCI Exhaust Vac Ekr 154K 5f 3 7/4.7
..,a PK;~.
4 w;
k.
9
~.
Table 4.7 1 thnticello Containment Penetrations (continued)
Applicable Appendix IfREN RARRIFR Otfrt.N BARRTER Designation Description J Type Test Designation Type h stable Designation D
Testable Penetration AO-2379 10 Yes X-218 Torus-Reactor Building C
Vacuum Bmakers and DWV-8-2 14 Yes Ventilation Supply AO-2380 10 Yes DWV-8-1 14 Yes
~
t X-219 RCIC Exhaust Vac BRr N0fE (note 11)
CV-3311 11 Yes X-220 Oxygen Analyzer C
Sample Point CV-3312 11 Yes HPCI-9 5
Yes X-221 HPCI Turbine Exhaust C
l l
HPCI-14 5
X-222 HPCI Steam Line Draine NONE (Note 9)
HPCI-15 5
RCIC-16 5
X-223 RCIC Steam Line Drains NONE (Note 9)
RCIC-17 5
M MO-1987 h
y X-22hA RER B Suction NONE (Note 9)
MO-1986 4
5 X-224B RHR A Suction NONE (Note 9) y h2M1 4
X-225 HPCI Suetion NONE (Note 9)
MO-2062 4
15ut 37/4.7
~ ~ '
e
?.-
.~
Table'4.7.1 Monticello Containment Penetrations (continued) r l
Penetration Applicable Appendix INNER BARRIER OUTER BARRIER Designation Description J Type h st Designation Type h stable Designation
' Wpe
' Testable r
X-226A Core Spray _ B Suction NONE (Note 9)
MD-1742 4'
X-226B ~
Core Spray A Suction NONE ' (Note 9)
M0-1741 4
X-227 RCIC Suction 2100 4
NONE (Note 9) m X-228 Not Assigned X-229A, Spare Penetrations.
NONE 17 X-229c - 229K X-229B Instrument Air to Toruc C
CV-7956 9
No X-230 Electrical Penetration B
11 Yes-l l
l q[
i 1 '
I 1.,f i
l i
?
g.'
[
m 37/h.7 154M
, - n,.
i s
d:
~
' Table' h.71 - Monticello Containment Penetrations (continued) f Explanation of Notes:
1.
During refueling outages,.when the vessel head is removed, MSIV's are tested by pressurization between valves. Since. test pressure tends to unseat the inboani valve, a lower test pressure than Pa is specified.
~
. 2.
Isolation is accomplished using manual valves in the containment supply line. 1hese valves are opened only.
when containment integrity is not required. Ibe valves are closed in accordance with valve lineup checklists which are completed prior to plant heatup.
3 One-inch instrumentation lines equipped with excess flow check valves." Subject to leakage testing in accordance with Technical Specification 4.7.D.1.b.
Icakage can occur only through rupture of the
~
line or its associated instrument 'outside of cont;ainment.
4.
TIP probes are withdrawn on a containment isolation signal and the line is isolated by automatic closure of a ball valve.. A shear valve can be manually actuated from the Control Room in the event a probe fails to retract. A solenoid valve in the purge supply line automatically closes on a containment isolation signal.
5 Containment isolation of the CRD hydraulic control lines is accomplished with a ball check valve internal to each drive mechanism and the nomally closed hydraulic system control valve.
6.
The drywell air lock is constructed with both doors opening inward so that containment pressure vill tend to seat the door seals. During overall air lock pressure tests, a st pport member is installed on the inner door to prevent the door fmm being forced open.
i 7
These are internal penetrations between sections of the containment structure.
g.
s 8.
Instrumentation lines not equipped with excess flow check valves. Leakage can occur only through rupture of l
the line or its associated instrument outside of containment.
I 9
This penettstion terminates at the bottom of the suppression pool.
It is not exposed to the containment atmosphere.
7
- 10. These drains are installed at the bottom of the suppression pool.
UN
- 11. Ibe HPCI and RCIC steam exhaust line vacuum breaker penetrations utilize the HPCI and RCIC steam exhaust l
line check valves'for containment isolation.
l 15kN i
37/4.7 REV
.h $N
.i
.~
Able 4.7.1. Monticello Containment Penetrations (continued)
Barrier Type Codes 1
Double gasketed seal 2
Hot pipe expansion bellcus 3
Air operated globe valve 4
Motor operated gate valve 5
Testable check valve 9,.
7 Air operated gate valve
'J 8
Manual gate valve 9
Diaphragm air operated control valve 10 Air operated butterfly valve
~
11 Electrical penetration 12 Motor operated globe valve 13 Manna 11y operated globe valve 14 Self-actuating vacuum breaker 15 Ball Valve 16 Solenoid Valve 17 Spare Penetration - velded cap 18 Instrtunent Line with excess flow check valve 19 Instrument Line without excess flow check valve
%,s I
w f
5 l
id -
I l
37/4.7 154 0
.e
j$
W,,
Bases Continued:
rate and a. standby gas treatment ' system filter efficiency of 90% ' for halogens, 9% for particu-lates, and assuming the fission product release fractions stated-in TID 14844, the calculated maximum total whole body passing cloud ' dose is about 3 rem and the calculatid maximum total thyroid dose is'approximately 150 rem at the low population zone distance of one mile for the duration of the accident. The resultant doses that were calculated for the two-hour accident
- dose -at' the exclusion zone boundary of'1600 feet ' are lower than the above. stated doses. Thus,.
,E.
the doses reported are the maximum that would be expected in the unlikely event of a design.
L-basis loss of coolant accident. These doses are also based on the assumption of no holdup in the' secondary containment resulting in a direct release of fission products'from the primary containment through the filters and stack to the environs. Therefore, thei specified primary containment leak rate and filter efficiency are conservative and provide margin between
~
off-site doses and 10 CFR 100 guidelines. The fission product source term defined in was also used in the design of the facility engineered safety features, including filter sizing.
The maximum allowable leakage mte, Ie,.at the calculated peak accident pressure Pa of 41 psig at test conditions was derived from the' maximum allowable accident leak rate of about 1. % per day when corrected for the effects of containment environment under accident and test conditions.
In the accident case, the containment atmosphere initially would be composed of steem and hot air depleted of oxygen, whereas under test conditions the test medium would be air or nitrogen at ambient conditions. Considering the differences in mixture composition and temperatures,-
the appropriate correction factor applied was O.8 as detennined from the guide on containment testing and results in an La of 1.2 wt%/ day. (4)
(4) TID 20583, Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Rate Determinations y
v' h.7 BASES 162
'O REV ms y w.
~~
I E
1
?
l l
Bases Continued:
Although the dose calculations suggest that the accident leak rate could be allowed to increase to about 2.4% per day before the guideline thyroid dose value given in 10 CFR 103 would be exceeded, establishing the test limit of 1.2% per day provides an sdequate margin of safety to assure the health and safety of the general public. It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Additional
,6 margin to maintain the containment in the "as buJlt" condition is achieved by establishing the allowable operational leak rate. The operational limit is derived by multiplying the allowable test leak rate by 0.75, thereby providing a 25% reargin to allow for leakage deterioration which may occur during the period between leak rate tests.
I The surveillance testing requirements for overall integrated contaiment leakage tests (Type A l
tests) and local penetration leakage tests (Type B and Type 'C tests) are consistent with the "Prf mary Reactor Containment Ieakage Testing for Water-requirements of 10CFRgO, Appendix Jwith the following exceptions:
Cooled Power Reactors
- 1. The test pressure for the air lock and the testing required following each use departs from Appendix J because of special design considerations.
The leakage limitations and surveillance requirements provide assurance, I
however, that this component is leak tight while at the same time providing a reasonable degree of freedom of access to the containment.
l
- 2. Main steam isolation valves must often be tested by pressurizing the connecting volume between the two valves. The maximum air pressure which can be used in this test is 25 psig.
Bis departs from the Appendix J requirement for f
Type C tests at kl psig. Testing at 25 psig has been shown to p mduce egnally valid leakage measurements, however.
w
- 3. Appendix J can be construed to require testing of a number of components which y
lack the pmvision for. Type B or Type C testing. 'Ihese exceptions to Appendix J requirements.are noted in Table h.71.
Icakage from these penetrations following an accident is improbable.
4.7 BASES 163 PJLV
-nn
..w j
fl t
.so 4
e Bases Continuad:
1 I
~
- 4. In a nunber of cases isolation valves are tested with pressure applied on the opposite side of the valve than the side pressurized followir4 an accident. This is done where tio provision exists for testing the valve l
in the correct direction and generally results in a more conservative leakage measurement.
Results of loss of coolant accident analyses indicate that fission products would not be released directly te the environs because of leakage through the main line isolation valves due te holdup in the steam system complex. Although this effect shows that en adequate margin exists with regard to release of fission products, the results of leak tests on the main steam line isolation valves will be closely followed in order to determine the adequacy of these valven to perform their intended function.
h i
1N 4.7 BASES g
-- W #in*1w % -
.n
uncionu gg
~
u.s. Nucuan nicutatony MISSION DOCKE T NUMut n IM5' 50-263 NRC CISTRIBUTION ron PART 50 DOCKET MATERIAL "To:
FHOM: NSP DATE oF DOCUMENT Mr. V. Stello Minneapolis, Minn.
55401 5-4-76 L.O. Mayer DATE HECEIVEo 5-12-76 L3tt TTc.:
.O NoToniz c o enor iNrurronM Nuusen or corits sectivto OonsciNAL
[3vscLAssiric facory 40 signed
~
Ltr re their 9-19-75 ltr 6 their 1-30" tNetosvac Revision to ltr dated 9-19-75 from osscnirTion 76 submi teal....trans the following:
NSP to NRC re Monticello Compliance with the Requirements of 10CFR50, App.
J....
Revision to Lic. Amdt Request dated 1-30-76 from NSP to NRC with attached revised Exhibits A, 6 B.....
.t (40 cys ea 'enci ree'd)
AC R 0W W3$}
P N NAME:
M ntieello Pla,nt t
SAFETY FOR ACTION /INFORMATION ENVIRO DHL 5-13-76
'I ASSIGNED AD :
ASSIGNED AD :
Ig liCGiCH CHIEF :
[g ) ' 1i e M NAl BRANCil CIIIEF :
i at PROJECT MANAGER:
fgdg PROJECT MANAGER :
- C t/ LIC. ASST. :
Q1ggs LIC. ASST. :
t INTERN AL DIST RIBUTION
!A M EC FILE 7 SYSTEMS E EXTY I PLANT RYSTEMS FMVTnn TFru "i
NRC PDR-HEINEMAN TEnERCO ERI ST I-l# -I_& E [2}
SCMROFDER BENAROYA BALLARD l t#
OELD IAlHAS SPANGLER
'W COSSICK & STAFF ENGl? JEERING IPPOLITO FilEC MACCARY SITE TECil CAsg KNIGHT OPERATING REACTORS GAF011LL HAHAUER SIINEIL STELLO STEPP _
MAgtESS PAWLICKI HUDIAN t
i OPERATING TECl!
l PROJECT MANAGEMENT REACTOR SAFETY
/
EISENillTr SITE ANALYSIS BOYD ROSS j SilAO VOLUIER P. COLLINS NOVAK V BAER BUNCH
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1100STON ROSZTOCZY v
SCIIWENCER M
J. COLLINS l_
PETERSON CilECK
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CRIMES KREGER MiiLTZ
,I llELTEMES AT & I SITE SAFETY & ENVIEC SKOVn0LT SALTZMAN ANALYSIS RiffBERG DENTON & MULI.ER EXTERNAL 0101 HIDUTION COff1Tb3L NUMBEH M___LPDIU/Ma'N N F# A h*$ f}L TL LAB liR00KilAVl'.N NATL LAB 1
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