ML20092G600
| ML20092G600 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/22/1984 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| 1054, GL-83-28, TAC-53170, NUDOCS 8406250164 | |
| Download: ML20092G600 (17) | |
Text
s e
e Docket No. 50-346 TOLEDO
%mm EDISON License No. NPF-3 Roano P. CnouSE Serial No. 1054 vc. pm.t wcw June 22, 1984 Director of Nuclear Reactor Regulation Attention:
Mr. John F. Stoiz Operating Reactor Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission Washington, D. C.
20555
Dear Mr. Stolz:
This is in response to the Safety Evaluation Report (SER) dated April 30, 1984 (Log No. 1503) which addressed the proposed modifications tc the Toledo Edison's Davis-Besse Nuclear Power Station, Unit 1, Reactor Trip System. These modifications were developed in response to the Rr; quired Actions Based on Generic Implications of Salem ATWS Event (Generic Letter 83-28, Log No. 1322 dated July 8, 1983).
In the SER, the staff concluded that certain aspects of the originally proposed design were inadequate. Toledo Edison has modified its Reactor Trip System design and has included the details of this design in the attachments. Within this submittal, responses to each specific concern identified within the referenced SER are included.
TED concludes that the mqdified TED design is consistent with the B&W Owners position, and is similar to the AP&L design for ANO-1 nuclear unit which is now approved by the NRC. Additionally, the modified TED design satisfies the conditions for approval set forth in the conclusion of the SER (Lcg No. 1503).
Very truly yours, 0 P W/+
RPC:JSH:nif enclosures cc: DB-1 NRC Resident Inspector 8406250164 840622 PDR ADOCK 05000346 P
PDR THE TOLEDO EDISON COMPANY EDISOO PLAZA 300 MADISON AVENUE TOLEDO, CJO 43652 b!
- Dock;t No. 50-346 Lic nsa No. NPF.
Serial No. 1054 Attachment I' Page 1 REACTOR TRIP BREAKER SHUNT TRIP CIRCUIT MODIFICATION l
DAVIS-BESSE UNIT NO. 1 TABLE OF CONTENTS I
Attachment I Reactor Trip Breaker Shunt Trip Circuit Modifi-cation Davis-Besse Unit No. 1 (Pages 1 thru 4)
Attachment II Test outline (Page 1)
Attachment III-DBNPS Technical Specification (References)
Sketch No. 26372-1 Reactor Trip Breaker Arrangement Sketch No. 26370-2 Schematic Diagram Reactor Trip Circuit Breaker (A & B)
Sketch No. 26371-2 Schematic Diagram Reactor Trip Circuit Breaker (C & D) f i
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' Dock:t Na~.~50-346 Lictase Na. NPF-3 Serial No. 1054-Attachment I Page 2 REACTOR TRIP BREAKER SHUNT TRIP CIRCUIT MODIFICATION DAVIS-BESSE UNIT NO. I 1
BACKGROUND By letter dated December 9,1983 (No.1012)) Toledo Edison Company (TED) submitted a-response to Item 4.3 of Generic Letter 83-28.
The submittal.
included a description of the TED's design of shunt trip modification (TED's design) for the reactor trip breakers (RTBs) at Davis-Besse Nuclear i
Power Station (DBNPS) Unit No. 1.
Also, described were the differences of
. Arkansas Light and Power-Company's (AP&L) design for Arkansas Nuclear One Unit No. 1.(ANO-1) and TED's design for DBNPS Unit No. 1.
In addition, TED's submittal included responses to specific. questions identified in the NRC evaluation of the ANO-1 design which.was provided to Toledo Edison Company by the NRC letter of September 22, 1983.
Subsequent to the review of TED's submittal, staff issued, by letter dated April 30, 1984 ( 503), a Safety Evaluation Report (SER), indicating the acceptable and oaacceptable aspects of the TED's design. The staff requested that 12D submit the modified design along with the proposed technical specification for the shunt trip by June 22, 1984.
-As indicated in the SER,' approval of the modified design would be con-ditioned on TED's ' response to Items (a) thru (f). described in the Con-
.clusion (Pages 8 thru 9) of the SEh.
Toledo Edison has modified the design for the addition of shunt trip' devices, on the reactor trip breakers, to be consistent with the B&W Owners Group's position and to answer the concerns indicated in the SER.
PURPOSE.
This letter describe's the TED's modified-design and, addresses Items ~(a) thru (f) of-the SER conclusions. TED intends to' implement this modification during the 1984 refueling outage.
MODIFIED' DESIGN-The TED's modified _ design of the shunt t' rip modification is same as:the AP&L. design for AC breakers.. The specificsLof.the TED's design are as follows:
UV Sensor
- (Sketches Nos. 26370-2, 26371-2)
.In the modified design the Model.ITE-27H-211R' relay, operable 1from a Class
.IE 125 VDC power ~ source, is used to actuate the. shunt trip attachment.'
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. Docket No. 50-346
- Liccate No. NPF-3 Serial No.-1054 Attachment I Page.3 LTED will assure that the seismic qualification of this relay envelops the design basis seismic conditions of the location of RTBs at DBNPS Unit No. 1.
Power Supply (Sketch No. 26372-1)
Power to shunt trip attachment of each reactor trip breaker (RTB) will be provided from a separate Class 1E power source. The cable carrying the control power will be Class 1E qualified and the associated conduit run will be installed to Seismic I. category.
Control power to the non-safety related source interruption device (SID) will remain from the existing non-safety related 125 VDC source.
Isolation between the non-safety related and the safety related wiring has been provided within the breakers and the cable routing outside the breakers.
+
I Shunt Trip Attachments (Sketch Nos. 26370-2,26371-2) j In the modified design, the shunt trip attachments will be designated as j
safety related. The shunt trip attachments in all four breakers will be seismically qualified, based on the qualification testing of similar i
attachments in AK-2 breakers of Unit Electric Control, Inc. We will assare that the test results in this report are acceptable according to the_ industry standards and applicable to the RTBs at DBNPS Unit No. 1.
Indicating Lights (Sketch Nos. 26370-2,26371-2)
F Two indicating lights are provided.in the modified design.- One will be lit when power is available for the shunt trip ' attachment, and the second will be lit when power'is available for the under voltage sensor.
RESPONSE TO ITEMS (a)'THRU (f) IN THE CONCLUSION OF SER ON PREVIOUSLY SUBMITTED TED DESIGN Item-(a) Confirmation ~of the seismic qualification of the UV sensor-(ITE-27H-211B).
Response
The modified TED design no-longer contains this' relay. 'The new
. relay is Model'ITE-27H-211R, which is same model as in the AP&L design. We will assure, prior to use, that the seismic:qualifi--
- cation of this-relay envelops-the RTB's seismic requirements at Lthe DBNPS Unit No. 1.
_ Item-(b). Confirmation (th'at the breakers with AC shuntitrip. coils are-seismically qualified.
4 ll:
Docict No.'50-346 Lic:nsa No. NPF-3
. Serial Fo.'1054
' Attachment,I Page-4
Response
TED's modified design no longer uses AC shunt trip coils, therefore this question is not applicable. The DC coils used will be verified to be seismically qualified, prior to use, for Davis-Besse Station.
Item (c) Designation of the automatic shunt-trip circuits as safety-related and incorporation of design features as defined in Items 4 and 5 above.
Response
l The shunt trip attachments are now designated as safety related.
Both undervoltage and shunt trip circuits of each RTB are powered from the separate Class 1E sources (120 VAC and 125 VDC respectively), thus maintaining the channel separation among the RTBs.
There is channel separation in the cable routing of the UVD trip
' circuits, and the same be provided in the routing of cables for shunt trip circuits.
Isolation of non-safety related SID from the IE shunt-trip will
-be provided through the coil to contact isolation of a Class IE relay (94).
(see drawing 26370-2) l A result of the modified design is that the statements made in Section 4b of the SER are not applicable.
Item (d) Incorporation of status indicating lights in the design and sub-mission of revised test procedures as defined in' Item 6 above.
Response
The additional status' indicating lights are included in the modified design. The test outline is included as Attachment'II to this letter.
Item (e) Submission of revised technical specification as defined in Item 7 above.
Response
-Davis-Besse Unit'l is covered by standardized technical specifica-tions. Attachment III-contains copies:of the' applicable portions relating to Reactor Trip Breakers. Section 4.3.1.1.1 requires-that surveillance testing be accomplished-by performing a CHANNEL FUNCTIONAL TEST of the Control Rod Drive Trip Breaker on
' Dock:t No. 50-346 Licznse No. NPF-3
- -Serial No. 1054 Attachment I Page 5 a monthly basis. The definition of CHANNEL FUNCTIONAL TEST (Attachment III, Page 1-3) includes verification of alarm and/or trip functions.
TED interprets trip functions to include all trip functions which encompasses the shunt trip attachment to reactor trip breakers. The test of the shunt trip attachment can only be done independent of the test of the UVD to assure its function.
TED's surveillance test procedures will be modified to include testing of the shunt trip attachments when they are installed.
We therefore conclude that the existing technical specifications explicitly state that testing independently confirms the oper-ability of both the shunt and undervoltage trip attachments.
Item (f) Submission of revised electrical schematics to reflect the changes defined in Items 4, 5 and 6 above.
Response
Revised electrical schematics Nos. 26370-2, 26371-2 and 26372-1 are attached to this letter.
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~ Dockst=Nn. 50-346 Licensa No. NPF-3 Serial No. 1054
-Attachment II Page 1 TEST OUTLINE r
l
1.0 OBJECTIVE
4.
The purpose of this outline _is to list the sequence of operations required to independently _ verify the operability of the UVD and shunt trip attachment of the four reactor trip breakers A, B, C & D.
It is being assumed that Anticipatory Reactor Trip' System (ARTS) will be in normal functional state at the time of this testing.
2.0 PROCEDURE
(Refer to Sketches Nos. 26370-2 and 26371-2) 2.1 Testing of UV trip circuit and the alarm relay (typical for all four breakers).
2.1.1 Verify that the breaker under test is in closed position and the control power to the shunt trip circuit of the breaker is ON (DC indicating light on i
the breaker is lit). Also verify that Reactor Pro-tection Systet (RPS) is energized and the AC indi-cating' light on the breaker is lit.
2.1.2 Turn the key operated hand switch to "UV Test"
. position and hold.
2.1.3 Observe that the DC indicating light goes out and confirm that an alarm is received in the Control Room indicating a loss of shunt trip circuit control power.
Also observe that the AC indicating light remains lit.
2.1 ~ 4 At the Reactor Trip Module of the RPS, place the 4
Reactor Trip Module Reactor Protection channel-switches "A" and "B" in the "Sim Trip" position. -This will actuate the relay Contacts A and B in the under--
voltage trip circuit of the breaker associated with-the RPS channel actuated.
2.1.5 Verify breaker tripping by observing the breaker local indication'and the AC indicating light going out, instanteously.
12. 1. 6.
Obse'rve_that upon releasing the test switch knob,~the IX:. indicating light is' lit again indicating resumption of control' power to the shunt trip ~and the AC indicating light remains out.
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' Docket No. 50-346'
-Lic nsa No. NPF-3 Serial No. 1054 Attachment II Page 2 2.2. Testing of the shunt trip circuit (typical for all four breakers).
2.2.1 Verify that the breaker under test is in closed position and the control power to the shunt trip circuit is available (DC indicating light is lit).
Observe that the AC indicating light is also lit indicating the RPS energized.
2.2.2 Turn the key operated hand switch to " shunt trip" position, the breaker will trip and the two indicating lights go out.
3.0 ACCEPTANCE CRITERIA:
Each Reactor Trip Breaker trips ~ normally, both on UV and shunt trip tests.
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IAOLI 3 3*1 License NO. %FF-3 seris.1 l0.1051 REACTOA PROTECTION SYSTEM INSTRtMENTATI0h
- , Attar.narnt III g
ras 1.
g TOTAL NO.
CHANNELS CHANNELS APPLICABLE
~ g, 0F CHANNEL 5 TO TRIP _.
OPERABLE MODES ACTION g
FUNCTIONAL UNIT-1.
Manual Reactor Trtp 2
1 2
1, 2 and
- 1 X
e 4
2 3
1, 2 2f
.5 2.. High Flus 4
2 3
1, 2 3#
-e 3.
RC High Teg erature 4.
Flux - a Flux - Flow 4
2(a)(b) 3 1, 2 2f 4
2(a) 3 1, 2 3#
5.
RC Low Pressure 4
2 3
1, 2 38 6.
RC High Pressure 4
2(a) 3 1, 2 38 7.
RC Preswn-Temocrature 8.
High Flux /Naber of Reactor Coolant 4
2(a)(b) 3 1, 2 3#
Pugs On 4
2 3
1, 2 3#
9.
Containment High Pressure
.t:'
- 10. Intennedtate Range, Neutron Flux 2
0 2(c) 1, 2 and
- 4 and Rate m
4,
- 11. Source Range, Neutron Flux and Rate 2
0 2
2ff and
- 5 A.
Startup B.
Shutdown 2
0 1
3, 4 and 5 6
- 12. Control Rod Drive Trip Breakers 2 per trip 1 per trip 2 per I, 2 and
- 7#
system system trip system
- 13. Reactor Trip Module 2 per trip 1 per trip 2 per 1, I and
- 7f system syste:s trfp system 3"
6f 2 ".., $..
- 14. Shutdown Bypass High Pressure 4
2 3
4
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TABLE 3.3-1 (Continued)
TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
TABLE NOTATION evith the control rod drive trip breakers in the closed position and and the inoperable channel above may be bypassed for the control rod drive system capable of rod withdrawal.
up to 30 minutes in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period when necessary to test the trip brnaker associated with the logic of the channel being tested per Specificat, ion
- Men $hutdown Sypass is SCtuated.
4.3.1.1.1, and
- The provisions of Specification 3.0.4 are not applicable.
c.
Either. THERMAL POWER is restricted to
- 75% of RATED RATED THERMAL and the High Flux Trip 5etpoint NNigh voltage to detector may be de-energized above 10-10 anos on both is reduced ta < 85% of RATED THERMAL POWER within 4
.Intevisediate Range channels.
hours or the QUADRANT POWER TILT is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(a) Trip may be manually bypassed when RCS pressure < 1820 psig by actuating Shutdown typass provided that:
ACTION 3 - With the number of OPERA 8LE channels one less than the Total Number of Channels STARTUP and POWER OPERATION mav i
(1) The High Flux Trip 5etpoint is 1 s of RATED THERMAL proceed provided both of the following conditions are '
E
- POWR, satisfied:
(2) The Shutdown Bypass High Pressure Trip $etpoint of i 1820 t
a.
The inoperable channel is placed in the tripped psig is imposed, and condition within one hour.
(3) The Snutdown Bypass is removed when RCS pressure > 1620 psig.
j b.
The Minisun Channels OPERABLE requiren.ent is j
met; however, one additional channel may be bypassed
(
(b) Trip may be manually bypassed when Specification 3.10.3 is in for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per effect.
Specification 4.3.1.1.1 and the inoperable channel above may be bypassed for up to 30 minutes in any 24 (c) The minimum channels OPERA 8LE requirement may be reduced to one hour period when necessary to test the trip breaker associated with the logic of the channel being when specification 3.10.1 or 3.10.2 is in effect.
i tested per Specification 4.3.1.1.1.
ACTION $TATEMENTS ACTION 4
- With the number of channels OPERABLE one less than re-With the number of channels OPERABLE one less than required quired by the Minimum Charmels OPERABLE requirement and ACTION 1 by the Minimum Channels OPERABLE requirement, restore the with the THERMAL Power level:
inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least NOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or a.
< 5% of RATED THERMAL POWER restore the inoperable open the control rod drive trip breakers.
channel to OPERABLE status prior to increasing THERMAL POWER above SE of RATED THERMAL POWER.
ACTION 2 - _ With the number of OPERABLE channels one less than the Total Musber of Channels STARTUP and/or POWER OPERATION b.
> 5% of RATED THERMAL POWER, POWER OFERATION may may proceed provided all of the following conditions are continue.
satisfied:
a.
The inoperable channel is placed in the tripped condition within one hour.
i b.
The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per
$pecification 4.3.1.1.1
. DAVIS-BESSE UNIT 1 m
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TABLE 3.3-1 (Continued)
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ACTION STATEMENTS (Continued)
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With the num6er of chonnels OPERABLE one less than required
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above 10-amps on the In instrume
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ACTION 6 With the nuber s'f channels OPERABLE one less then re-
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m verify compliance with the 5MJT00WM RARGIN requirements
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mlT of Specification 3.1.1.1 within one leur and at leest once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf ter, 3^
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ACTION 7
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k" Total Number of Channels 51ARTUP and/or POWER OPERATION With the number of OPERABLE channels one less than the gg
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R F 2 5
may proceed provided all of the following conditions are satisfied:
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e.
Within I hour:
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Place the inoperable channel in the tripped
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device associated with the inoperative channel.
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5 5
hours for survettlance testing per Specification 4
I"T
.g 4.3.1.1.1, and the inoperable channel above any be 4
bypassed for to 30 minutes in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period
-eh" when necessary to test the trip breaker associated i'
- E with the logic of the channel bging tested per 5pecification 4.3.1.1.1.
The 16 operable channel above T E" sty net be bypassed to test the Ingle of a channel of the trip system associated with the inoperable g
channel.
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ACTION 8 With the neber of channels OPERABLE.less then required by the Minfam Channels OPERABLE requireent, be in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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TABLE 4.31 BfACT04 PROTICTION 5751 01 INSTalmINTATION SURVtILLANCI afdlletMfuTS G
M 04AlelfL 20(! IN WICM E
CNAustL CNamart rmCTionAL SuavtILLAnCC "e
FUNCTIONAL Uulf CHECE CAlltAAfilpl TEST RfQUIRED 1.
Menuel Beector Trip R.A.
E.A.
5/U(1)
N.A.
2.
Nigh Flus S
e(2).andg(r)
M
- 1. 2 3.
RC Nigh Temperature 5
t N
- 1. #
4.
Flea - AFles - Fles-S(4) 10(3)eedg(F.8)
M
- 1. 't S.
RC Les Pressure l'
t***
M
- 1. 2 8
aC Nigh Pressure S
B***
N
- 1. I g
7.
AC Pressure.Tamperatore 5
R R
- 1. 2 8.
Nigh Flus /Humber of Beecter
{
Caeleet Pimps to S
S N
- 1. 2 1.
f.antalmient Nigh Pressure.
5 E
N
- 1. I le. Inter.ediate sene.. Neutree fles e ( Rete S
a(7)
$/U(lX1)
- 1. 2 and*
- 11. Seerce Seege, asutree rios 4
- E{si C
- =.d1/v( Q M "Me 12 Centrol Red Drive Trip Breaters B.A.
N.A.
Mand 5/U(1)
- 1. 2 and*
g Tripnee.fo45s
%A4.
T
^
ada 14 Shutdoen typass Nigh Pressure S
R N
2 **.3**.4'* $
c
.I'krCF 1:.;lj &
. 4:.c
- .8 d I
EE V "E
DEFINIT!0ms 1.0 DEFINITIONS CHAmEL FtseCTIOmAL TEST DEFINED TERMS 1.11 A CHANNEL FlseCTIONAL TEST shall be:
1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical $pecifications.
Analog channels - the injection of a simulated signal into the a.
chsanel as close to the primary sensor as practicable to verify OPERASILITY including alarm and/or trip functions.
THERMAL p0WER b.
Sistable channels - the injection of a simulated signal into 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to ser to verify OPERASILITV including alarm and/or the reactor coolant.
r RATED THERMAL p0WER CORE ALTERATION 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate 1.12 C0er ALTERATION shall be the movement or manipulation of any com-to the reactor coolant of 2772 fatt.
ponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.
OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall cor espond to any one inclusive comoina-SMUTD0wh MARGIN
(-
tion of core reactivity condition, power level and everage reactor coolant temperature specified in Table 1.1.
1.13 SMUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or mould be suberitical fram its present condition assuming:
ACTION a.
No change in antal power shaping rod position, and 1.5 ACTION shall be those additional requirements specified as corollary statements to each principle specification and shall be part of the b.
All control rod assemblies (safety and regulating) are fully specifications.
inserted except for the single rod assembly of highest reactivity 8
morth stich is assumed to be fully withdrawn.
,q 1
p OPERABLE - OPERA 8!LITY IDENTIFIED LEAKAmT 1.6 A system, subsystem, train, corponent or device shall be OPERABLE or have OMRASILITY when it is capable of performing its specified function (s).1 1.14 IDENTIFIED LEAKAGE shall be:
Imp 11 cit in this definition shall be the assumption that all necessary
[
attendant instrumentation, controls, normal and emergency electrical power Leakage (except CONTROLLED LEAKAGE) into closed systems, such sources, cooling or seal water. lubrication or other auxiliary equipment, a.
as pump seal or valve packing leaks that are captured and that are required for the system, subsystem, train component or device conducted to a swap or collecting tant. or to perfom its function (s), are also capable of performing their related support function (s),
b.
Leekage into the containment atmosphere from sources that are w th the o eek t
ens or t
be PRESSURE BOUNDARY LEAKACLE. or D
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DEFINITIONS c.
Reactor coolant system leakage through a steam generator to aEPORTABLE OCCURRENCE the secondary system.
1.7 A REPORTABLE OCCUR *INCE shall be any of those conditions specified in Specifications 6.g.1.8 and 6.g.1.g.
tm!XCIFIED trarau i
1.15 mIDECIFIED LEAEAGE shall be all leakage seich is not IDECIFIED CO CA! WE C INTEGRTTY LEAKAE or CONTROLLED LEAKAGE.
N8 CONTAIMENT INTEGRITY shall exist when:
press;;c! BOUaCARY LEAKAGE a.
All penetrations required to be closed during accident con-ditions are either:
1.16 PRES 5URE Boum3ARY LEAEAR shall be leakage (except steam generator tube leatage) through a non-tsolable fault in a Reactor Coolant System 1.
Capable of being closed by the Safety Features Ac'tuation component body. pipe nell or vessel well.
System, or ConMcLLED LEAKAE I
2.
Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.
1.17 CDCR1 LED LIAEAGE shall be that seal water flow from the reactor f
except as provided in Table 3.6-2 of specification coolant piano seals.
(
3.6.3.1.
OUA3s u T p0WER Titt b.
All equipment hatches are closed and sealed.
C 1.18 QJAORA C POWER TILT is defined by the following equation and is c.
Each airlock is OPERABLE pursuant to Specification 3.6.1.3.
expressed in percent.
l d.
The contalment leakage rates are within the limits of Specification QUADRANT P0tER TILT =
3.6.1.2, and P
i
- I ower in a*y core quaerantAverage poner of ail quadrants 'N e.
The sealing mechanism associated with each penetration (e.g..
welds, bellows or 0-rings) is OPERASLE.
DCSE EQUIVALEwT I-131 CHANNEL CALIBRATION 1.19 DCSE EQUITALEXT I-131 shall be that concentration of I-131 hC1/ gram) 1.g A NR CALIEWIM shah be the adustment. as necessary. of the w'tica alone would produce the same thyroid ecse as the quantity and Channel output such that it responds with necessary range and accuracy isotopic mixture of I-131. I-132. I-133.1-134 and I-135 actually present.
to known values of the parameter which the channel monitors. The CHANNEL The tnymit dose conversion factors used for this calculation shall be CALIBR.M shah enemss the entire channel including the sensor and those listed in Table III of TID.14844.
- Calculation of Distance Factors aiare and/or trip fur.ctions, and shall include the CHAmEL FUNCTIOhAL TEST.
for Power and Test Reactor Sites.
CHANNEL CALIBRATION may be performed by any series of sequential, over-lapping or total channel steps such that the entire channel is calibrated.
[- AVERAE OISIWTEGIATION EkEDGY 1.20 T-AVERAE DISICEGRATION ENERGY shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant 1.10 A CHANkEL CHECK shall be the qualitative assessment of channel at the time of sangling) of the sum of the average beta and gama energies behavior during operation by observation. This determination shall include, whe,e possible, conearison of the channel indication and/or status with other indications and/or s'atus derived from independent v
instrument channels seasuring the same parameter.
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