ML19309F500
| ML19309F500 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 04/22/1980 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML19309F494 | List: |
| References | |
| NUDOCS 8004290506 | |
| Download: ML19309F500 (45) | |
Text
. -..
O
' # 'o o n.9. sos Commonwealth Edison ATTACHMENT 1 l
Zion Station Units 1 and 2 i
NRC Docket Nos. 50-295 and 50-304 Technical Specification Revisions and Basis
)
o ATI'ACEFNI' 1 Technical Specification Revisions and Basis Page Revision Basis iii Delete words "and Relief" from Item 3.7.1 of Index Reference to steam generator relief valves was
- listing, rerreved in previous license amendments (Amendments No.17 and 14).
Item 3.7.2 should read " Auxiliary Feedwater Pump Changes index to conform to text on Page 158.
Systen" and the associated page number should be 158.
Item 3.8.4 has been added.
Changes index to conform to text on Page 173.
Item 3.9.4 has "and Bypasses" added to " Main Steam Changes index to conform to text on Page 200.
Isolation Valves."
vii Figures 3.4-1 and 4.16-1 are added.
Changes index to conform to existing figures.
6 Delete from definition of " Quadrant Power Tilt" Specification 3.2.2.C.2 states that irere thermo-the second sentence, "If one excore... the couples shall be used to determine quadrant power average."
tilt in the event of an inoperable excore detector channel.
19 Change "part length rods" to "the full-length control Part length rods have been removed from Unit 1 and rods".
Unit 2 during previous refueling outages (this justification also applies to revised Pages 41, 42, 46b, 47c, 51, 52, 68a, 69a, 72, and 297 described below):
the statenent on Page 19, however, should not be deleted because it also applies to the full-lenght rods.
Page Revision Basis 36 Under PERMISSIVES:
P-6 isset relative to the reading from For P-6 change channel Calibration the intermediate range neutron channel from "R" to "N.A." and Channel and thus should only be calibrated and Function Test from "M" to "S/U (1)",
tested at the same frequency as the add remark "(1) Not required if intermediate range channel.
performed within the previous 7 days."
For P-7 change Channel Calibration There is no instrumentation associated from "R"
to "N.A."
with this relay and a channel calibration is not meaningful.
36 For P-8 and P-10 change Channel The instrumentation feeding this relay is Calibration from "R" to "N.A."
required to be calibrated under Items 2 and 3 on Page 35 of the Technical Specifice.tions.
40 In Specificatica 3.2.1.C.1 after the To clarify specification so that criticality words "always negative" reposition during low power physics tests as well as and place in parenthesis the words all other operating modes, is allowed only "except during low power physics above 500 F as requested by NRC Staff in tests."
discussions relative to removal of the cold rod drop test (Amendment No. 47 to DPR-39 and Amendment No. 44 to DPR-48).
41 Change, in Specification 3.2.1.D.3, Same as for Page 19.
"but one full... capability. " to
" full length control rods fully inserted, except for the predicted most reactive rod."
Page Revision Basis 42 Delete Specification 3.2.1.D.6 Same as for Page 19.
Delete the words "part length rods Same as for Page 19.
fully withdrawn" from Specification 3.2.2.A.3.
44 In Specification 3.2.1.H, change To conform to existing regulatory agency (2 places) " Atomic Energy Commission name.
to "NRC."
46b Delete the words "part length rods fully Same as for Page 19.
withdrawn" from Specification 3.2. 2. A.3.
47b Delete this page.
No longer used.
47c Convert to Page 47b and delete Same as for Page 19.
Specifications 3.2.2.A.9 and 4.2.2.A.9.
48,49 In Specification 3.2.2.B, Quadrant Power The use of the word " ratio" is redundant Tilt Limits, remove the word " ratio" in that the definition of " quadrant power after the words " quadrant power tilt."
tilt" already includes a ratio.
It appears 5 times.
49,50 On Page 49, complete Specification Editorial 3.2.2.c l.b by adding from top of Page 50 words " power for 3 loop operation."
Remove same words from top of Page 50.
Page Revision Basis 50 In Specifications 3.2.2.C.2 and To conform to the definition of " quadrant 4.2.2.C.2, change the word " balance" power tilt" and to correct a typographical to " tilt."
Also, in Specification omission.
I 3.2.2.C.3.b insert the word " Condition" before the word "immediately."
1 51 In Specification 3.2.3.A.1, delete the Same as for Page 19.
words "or part-length."
52 Delete Specifications 3.2.3.B.4 and Same as for Page 19.
4.2.3.B.4.
68a In the second sentence of third Same as for Page 19.
paragraph, delete the words "part length rods withdrawn from the core and with."
In the third paragraph, add the words These words were inadvertently left out
" error are necessary and indicated of Technical Specification change submitted deviation" after the words "excore to the NRC on February 2, 1979 concerning detector."
ECCS/LOCA reanalysis and subsequently approved on February 16. 1979 (Amendment No. 42 to DPR-39 and Amendment 39 to DPR-48).
69a In the third paragraph, second Same as for Page 19.
sentence, delete the words "without part length rods."
Also delete the third and fourth sentences be-ginning with "It is accomplished...
power plant demand."
Page Revision Basis 70 In the last paragraph, remove the words To conform to the Jefinition of " quadrant "to average" and "The excore current power tilt."
tilt is indicated." Also, add the words
" quadrant power."
71 Change words " power tilt, " " quadrant To conform to the definition of " quadrant tilt" and " maximum-to-average" to power tilt."
" quadrant power tilt" in 8 places.
72 In the first sentence of the fourth Same as for Page 19.
paragraph, delete the words "and part-length rods."
95 In Specification 4.3.3. A.1, remove the Radiation detectors RE-00llA and RE-0012A word " particulate. "
monitor gaseous as well as particulate activity.
96 In Specification 4.3.3.C., change the Correction of a typographical error.
referenced sections from "4.3.4.A and 4.3.4.B" to "4.3.3.A and 4.3.3.B."
156 In Specifications 3.7.1 and 4.7.1, Same as for Page iii delete the words "and Relief."
173A In Specification 4.8.4.B.3, change To conform to the minimum ECCS pump flows 265 GPM to 275 GPM and in Specifica-used in the current ECCS/LOCA analysis as tion 4.8.4.B.4 change 432 GPM to 400 approved by the NRC Staff (Amendment No. 42 GPM.
to DPR-39 and Amendment No. 39 to DPR-48).
Page Revision Basis 187 Add "IEST to RHR Pump Suction Valves" and These valves are required for proper operation of MOV-PH8700A&B" to RHR canponent list, Table 4.8-3 RHR system but were inadvertently left off of Table 4.8.3.
297 In Specification 5.3, delete the words "and eight Same as for Page 19.
part length."
300 Replace Section 6.1.B with the following:
Replaces reference to Figure 6.1.1 of Technical Speci-fications with reference to Topical Report CE-1-A "The corporate management which relates to the (Camonwealth Edison's Quality Assurance Manual).
operaation of this station is as shown in Figure 1-6 of CE-1-A 'Ibpical Report. % e Division Manager, Nuclear Stations has the responsibility for the Fire Protection Program."
300 Replace the first two (2) sentences of Section Replaces reference to Figure 6.1.2 of Technical 6.1.C with the following five (5) sentences.
Specifications with reference to 'Ibpical Report CE-1-A (Comonwealth Edison's Quality Assurance
"%e normal functional organization for operation Manual).
of the station shall be as shown in Figure 1-6 and Figure A of CE-1-A Topical Report. W e manning for the station shall be as shown on Figure 6.1.1.
The Operating Assistant Superintendent, Operating Engineer, Shift Engineers, and Shift Forerran shall have a senior operating license. The Fbel Handling Foreman has a limited Senior Operating License.
An Operating Engineer will be responsible for implernentation of the Fire Protection Program."
300 Change " Administrative Assistant" to " Administrative To conform to existing station organization, and Support Services Assistant Superintendent" in Section 6.1.D.
~
Pace Revision Basis 301 & Change Sections 6.1.G.l.a. and 6.1.G.l.b. (ll) to include To conform to existing corporate organization.
302A "Vice President of Nuclear Operations" and change "Vice President of Construction,..." to
" Executive Vice President of Construction,..."
302A Change Section 6.1.G.l.C to replace " Executive To conform to existing corporate organization.
Vice-President" with "Vice Chairman" and to replace
" General Superintendent of Production Systems Analysis" with "Vice President of Nuclear Operations."
308 In Fpecifications 6.2.B.1 and 6.2.B.2 (two places),
To conform to existing e ' tion organization, change " Maintenance Engineer" to " Maintenance Assistance Suoerintendent."
329 E Delete Figures 6.1-1 and 6.1-2.
The Corporate and Station Organizations are now 330 referenced to Figare 1--6 and Figure A of Topical Report CE-1-A (Connonwealth Edison's Quality Assurance Manual).
331 Revise figure number fran "6.1.3" to "6.1.1".
Also
'Ib conform to existing station organization and to change " Radiation Protection Man" to " Rad-Chan chronological listing of figures.
Technician".
ATP/CINIM 2 Zion Station Units 1 and 2 NRC Docket Mos. 50-295'and 50-304 Proposed Technical Specification Changes
% e following pages have been revised:
iii 42 51 95 301 vii 44 52 96 302 6
46b 68a 156 303 19 47b 69a 173A 308 36 48 70 187 329 40 49 71 297 330 41 50 72 300 331
% e following page has te n added:
302A We following page has been deleted:
47c 1
- able of Contents (Continued)
SURVEILLANCE LI"ITI?!G CONDITIO" FOP OPERATIO'I 7PoUIDPHEt1T PAGE 3.7 Ftean Generator Frergency Feat Roroval M7 156 Stean Line Fafoty Valvon 4.7.1 156 I3.7.2 3.7.2 Auxiliary reedwater Purp Systen e.7.2 l
159 Eases 3.o Trarcency Core Coolinc and Core coolinc Furnort 4.c 16M 3.".1 Centrifucal Charnina Purn Fvstem a.o.1 164 3.".2 Fafety Iniection Purp System 2.0.?
]FS 3.o.3 Residual Feat onmoval Purp Syster A.o.3 170 costinc of Centrifucal Charcinc, Fafety Syster 13.".4 In4ection, and Pesidual Fea t P.croval Purp."" ster s l 4.S.4 l
173 3.".5 Accunulator System 4.9.5 174 3.0.6 Cor.ponent Coolina System 4.R.6 175 3.9.7 Service Water System 4.8.7 17R 3.?.R I!ydrocen Control Systems 4.R.8 1RO 3.?.9 Ecuiprent for Fvaluatina Post LOCA d.P.9 Iod Bases 3.0 Containrent Isolation Systers 4.0 147 3.o.1 Isolation Valve Foa3 Pater System t.9.1 197 3.o.2 Penetration Pressurization Systens 4.0.2 loe 3.'.3 Containment Isolation Valves M.9.3 199 l 3.9.4 "ain Stean Isolation Valves and Pvnasses 4.9.4 200 3.9.3 Containment Intecrity 4.0.5 201 cy 3g; Bases b )
3.10 Containrent Structural Intoarity 4.10 212 ate Testing 2.10.1 212 (ar) 3.1^.1 Containrent Leakage c
?.16.2 Containnent Tender Testing 4.3n,2 pig gggg 3.19.3 End Anchoraaes and Adjacent Concrete Furfaces Irspection 4.~10.3 217
(,
9 3.16.4 Containment Liner Inspection 4.ln.a 21o 535k 3.19.5 Containrent Pressure 4.10.5 219 3.In.c Containment Terporature 2.16.F 219 bd :b nases
?.11 Padioactive Licuids a.21 2??
Eases 3.12 Endinactive cases 3.3?
23n
$5ED b
rases iii
LIST OF FIGURES (Continued)
Figure Page 3.3.2-1 Reactor Coolant System Heatup Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limitations 85 3.3.2-3 Effect of Fluence and Copper Content on Shift of 86 OF RT for Reactor Vessel Steels Exposed to 550 NDT
^
Temperature 3.3.2-4 Fluence at 1/4T and 3/4T as a Function of Full Power 87 Service Years 3.4-1 High Steam Line Flow Setpoint 131a 4.16-1 Location of Fixed Environmental' Radiological 278 Monitoring Stations 6.1.1 Zion Shift Manning Chart 331 vii
-J
c.
ouadrant Dover Tilt K.
onerable The cuadrant nower til t nor unit is defined Pronerly installed in the systen and as the ratio of the ncximun upner excore canable of nerforninc the intended detector current to the averace of the functions in the intended manner as unper excore detector currents or the verified bv testinc and tested at ratio of the naximum lower excore detector the frecuency recuired bv the current to the average of the lower excore Technical Snecifications.
detector currents whichever is creater.
L.
Oncratinn H.
Pated Therral Pover f
Performina the intended functions A steady-state reactor core output of in the intended nanner.
3250 MFt ner unit.
M.
oneratinc Cvcle I.
Reactor Pressure The interval between the end of one The pressure in the stean snace of maior re#ueling outano and the end a pressurizer.
of the next subsequent naior refuelinn outane oer unit.
J.
Refuelino Outace N.
Surveillance Interval Mhen Refueling Outace is used to
-desienate a surveillance interval ner Surveillance intervals, with the unit the surveillance will Fo norformed excention o# shi#t and dail" norinAq, durina the refueline outace or up to six are defined as the socci'ied nerio1 "g
ronths before the refueline outace.
plus or ninus 25t o' the soeci'ied b' 3 When a refuelina outace occurs within
- period, 8 ronths of the previous refueline (t.._;)
outace for a unit, the surveillance gggj testing need not be perforned.
The maxirun interval between surveillance b
0 tests is 20 months per unit.
M s e)
W E523 b
6
The curves are based on the folloving nuclear hot chanr.cl # actors: (2)
( = 1.55
{l+0.2 (1-P))
The hot channel # actors are also where P is the fraction of rated power, and on a DNB analysis as described in sufficiently larqe to account for the decree o# maloositioninc o# the Reference 3.
The c::pression for Pi!p-full-length control rods that is for o < n < l.0 renrecents the c##cct allowed before the reactor trio set on radial power shapes of the control coints are reduced and rod 'rithdraval rods at the insertion linits.
block and load runback r.av be recuired.
(5)
Rod withdrawal block and load runback occurs if reactor trio set noints are aporoached within a #ixed linit. (6)
R The Reactor Control and Protection 8
Sustom is desianed to orevent anv b
U anticinated combination of transient 6
9 M
i i
6 3
(2)
PSAR Annendix 3A cection 5.3 E352)
(3)
FFAR Apnendix 3^ 9ection 4.3 ba d (5)
PSAR Section 14.1.3 (6)
PSAR Section 7.2.2 b
19 4
Reactor Trip channel channel Channel r
Channel Descrintion chock Calibration Function Test Dg arks 17.
Low Stear Generator Level F
o in Coincidence with Feed Flow Stear Flow Mispatch 18 Low-Low Stear Generator Level S R
M
"(
}
l^.
Fafety Tniection
?;. 7.
'?. 7.
(1) vanual SI function chacP at U only 20 Turbire Trin
- . A.
7,.
r 21.
'ittoratic Reactor Trip Locic
'!. A.
N.A.
'! III (1) ggg ggy g Dreaker Oooninq DTRIFFIFFS 22.
P-6 U.7.
?'. A.
S /I' II I (1)
Vot required if perfor:ned within the orevious 7 days 23.
P-7
?!. A.
?!. A.
u
..A.
.u.
n_c
.e.
25.
F-10
'I. A.
".A.
d
!!o te :
Specified intervals may be adiusted +25S to accorrodate test schedules.
F - Once ner Shift D - Once per Day h9 M - Once per Month M
EPPM-Once Per Effective Full Power Month i
i O - Once per Ouarter h-c d) 1 S/U - Prior to Startup R.- Once per Defuelina Shutdovn - calibration of these instrurents ray he done as ruch as 7M ronths prior to the start of refueline outace and still se tisf,y this recui rerent.
)D
'!. A. - Not Applicable g
- Applies to Unit I and Unit II TABLE 4.1-1 (Sheet 2 of 2)
Reactor Protection System Testinc and Calibration Reouirererts 3F
O*
O O'
LIMITINd CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4.2.1.B once a shift while remaining in this con-dition.
During heatup, the boron concen-tration in the reactor coolant loops and pressurizer shall be sampled every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The reactor coolant loop boron concentra-tion must not decrease by more than 50 ppm between successive 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> samples.
The pressurizer boron concentration must not be more than 200 ppm less than the reactor coolant loop boron concentration.
C.
Unit Startup C.
Startup 1.
Immediatel'y prior to startup, the reactor i.
The Tavg of each reactor coolant loop coolant temperature shall be shown to be shall be logged before attempting to greater than the temperature above which bring a reactor critical.
the moderator temperature coefficient is always negative (except during low power physics tests) and greater than 500 *F.
2.
When a reactor is approaching criticality, 2.
Not Applicable.-
the shutdown banks shall be fully with-drawn in sequence (shutdown bank A,B,C,D) before any other rods are withdrawn.
The control group rods shall be no further in-serted than the limits shown by Figure 3.2-2 for Unit 1 and Figure 3. 2-4 for Unit 2 for 4-loop operation and Figure 3.2-3 for Unit 1 and Figure 3.2-5 for Unit 2 for 3-loop operation when critical-ity is attained.
D.
Power Operation D.
Power Operation 1.
When a reactor is critical, except for 1.
Rod operation shall be verified by physics tests and control rod exercises, partial movement of all rods every the shutdown rods shall be fully with-two weeks.
Rods which have been exercised within the i
i 40 j
)
)
)-
LIMITING CCNDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.1.D.l.
drawn and the control group rods 4.2.1.D.l.
past two weeks during normal shall be no further inserted than operation need not be verified the limits shown on Figure 3.2-2 Control rod bank positions with for Unit 1 and Figure 3.2-4 for respect to its insertion limit Unit 2 for 4-loop operation and shall be yerified once per Figure 3.2-3 for Unit 1 and shift.
Figure 3.2-5 for Unit 2 for 3-loop operation.
2.
Control bank insertion may be 2.
Control rod bank worths shall further restricted if the measured be measured following each control rod worth of all rods, refueling outage.
less the worth of the most reactive rod (worst case stuck rod), is less than the reactivity required to provide the design value of avail-able shutdown as shown in Figure 3.2-1.
3.
During physics testu and control 3.
Not applicable, rod exercises, the insertion limits need not be observed, but the lim-its in Figure 3.2-1 must be ob-served except during the low power physics test to determine total control rod worth and shutdown margin.
For this test the reactor may be critical with all. full length control rods fully inserted, except for the predicted most reactive rod.
4.
Three reactor coolant pumps per 4.
Prior to proceeding from hot unit shall be operating whenever shutdown to hot standby, verify a reactor is critical except that three reactor coolant pumps during natural circulation test, are operating except during f
(power - 8% full power) or low natural circulation tests or power physics testing.
low power physics testing.
41
)
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.1.D 5.
Reactor power shall not be increased 4.2.1.D 5.
Not Applicable above 60% of rated power with only three reactor coolant pumps in operation unless the overtemperature
~
AT trip setpoint and the P-8 inter-lock for three loop operation has been set in accordance with speci-I, fication 2.1.1.B.4.
E.
Rod Bank Assignment E.
Rod Bank Assignment Rod Bank Assignment shall be as Rod Bank Assignment shall be deline,ated in Figure 3.2-8.
Except verified after each refueling during physics tests, the sequence of outage, for the refueled unit.
withdrawal of the control banks, when going from zero to 100% power, is A, B,
C, D with control bank overlap.
F.
Boric Acid System (per unit)
F.
Boric Acid System (per unit) 1.
A reactor shall not be taken from 1.
Surveillance and testing of hot shutdown to hot standby unless the boric acid system shall the following conditions exist:
be performed as follows:
a.
One boric acid tank for that a.
Boric acid tank level, reactor contains at least 5140 concentration and tempera-gallons of 11.5% (but not great-ture shall be verified er than 13%) by weight boric prior to startup and acid solution at a temperature weekly thereafter.
of at least 145*F.
42
)
LIF.ITING' CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2.1 G.
Primary System Boron Concentration 4.2.1 G.
Primary System Boron Concentration Changes during Cold Shutdown Changes during Cold Shutdown When a boration or dilution oper-The opera' tion of at least one ation is in progress, at least one reactor coolant pump or one reactor coolant pump or one resi-residual heat removal loop shall dual heat removal loop shall be be verified before the start of operating.
a boration or dilution operation.
H.
Reactivity Anomalies H.
Reactivity Anomalies A normalization of the computed Reactivity anomaly evaluations boron concentration as a function shall be performed following of 'burnup shall be compared with the startups after shutdowns of 72 actual boron concentration of the hours or longer duration but coolant.
If the difference between shall not be required more than the observed and predicted steady-once if more than one such shut-state concentrations reaches the down occurs in a two month, period, equivalent of one percent in l
reactivity, the NRC shall be noti-fied within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and an evalu-ation as to the cause of.the dis-crepancy shall be,made and reported l
to the NRC within 30 days.
44
l LIF.ITING CONDITION FOR OPERATION SURVEILLANCE REQUIRDIENT
\\
3.2.2.A 4.2.2.A 2.2.c (Continued) 2.2.c.
(1) Power between the maximum and (2) A flux dif ference alarm shall minimum limits specified in indicate non-conformance with 3.2.2.A.2.2.a.
the 3% AI target band for BASE LOAD operation.
If the (2) AI within thc4 I target band alarm is temporarily out of as per Section 3.2.2.A.4 and service, conformance with the 3.2.2.A.5, except use 13% AI applicable limit and the flux target band instead of the
+6, difference shall be logged
-7% 6I target band.
hourly for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 2.2.d If any of the requirements of Section 3.2.2.A.2.2.c.
are not maintained then 2. 2,. d Not Applicable power must immediately be reduced to below the power limited by APDMS type 4.2.2.A.3 surveillance (Section 3. 2. 2. A. 2.1. ) and APDMS type surveillance must be The reference ecuilibrium indicated initiated if the power is above PT*
axial flux difference as a function of power level (called the target 3.2.2.A.3 flux difference) shall be determined The target flux dif ference at a given at least once per equivalent full power level, P is determined by O,
power quarter.
The target difference noting the indicated axial flux difference should be updated every effective at the power level with equilibrium full p wer month.
This may be done xenon conditions established in the using the measured value for that core and with the full length rod bank l
~more than 190 steps withdrawn.
P for month or by linear extrapolation using the two most recent measured the purpose of determining the ta get values.
The initial target flux value, should be as high a power level difference on a reload may be as practicable.
The target flux dif-determined from design predictions.
ference at any other level, P,
is equal multipled by to the target value of PO the ratio, P/P
- 46b -
O.
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3 2.2.A 8.
For the purpose of determining 4.2.2.A 8.
Not applicable, 0
penalties associated with deviations from the target band, time for use in applying Items 6.1 and 7.2 above shall be accumulated in the following maener:
8.1 For deviations at or below 50% power, time shall be accumulated such that a 1 minute actual deviation equals a 1/2 minute accumulative penalty in applying Items 6.1 and 7.2 above.
8.2 For deviations above 50%
power, time shall be accumulated in a 1 for 1 time basis in applying Items 6.1 and 7.2 above.
47b
SURVEILLANCE REQUIREMENT LIMITING CONDITION FOR OPERATION 3.2.2 B.
Quadrant Power Tilt Limits 4.2.2 B.
Quadrant Power Tilt Limits 1.
Quadrant. power tilt shall 1.
If an indicated quadrant g
be calculated and logged l
power tilt exceeds 1.02, except I
for physics tests, then within along with the individual 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, one of the following upper and lower excore cal-steps shall be taken:
ibrated outputs as follows:
a.
Once each shift at power a.
Correct the tilt, o r.
levels greater than 50%.
b.
Determine by measurement b.
Four times a shift and the core peaking factors and apply Specification following a load change of more than 10% power 3.2.2.A or at any power level above c.
Restrict core power level
, 50% if one or both quad-so as not to exceed full rant power tilt alarms rating less 2% for each are inoperable.
percent of quadrant power tilt beyond 1.0.
2.
Not Applicable.
2.
If an indicated quadrant power tilt exceeds 1.02 for a period l
of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without known "cause, or if sudden tilt re-occurs intermittently without known cause, the reactor shall be put in the Hot Shutdown Condition within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
How-ever, operation below 50% of rated power, for testing and/or correcting the tilt, shall be permitted.
48
LIMITIN3 CCNDITION FOR OPERATION I
SURVEILLANCE REQUIREMENT
}
)
3.2.2.B 3.
If an indicated quadrant power 4.2.2.B 3.
Mot Applicable tilt exceeds 1.09, except for physics testing, the reactor shall be put in the Hot Shutdown Con-dition; however, operation below SOS of rated power, for testing and/cr correcting the tilt, shall be permitted.
C.
Instrumentation (per unit)
?
C.
Instrumentation (per unit) 1.
Excore axial inbalance detector 1.
Excore axial imbalance system detector systen u.
The excore axial imbalance a.,,Not Applicable detector system shall be recalibrated at least every three effective full pover
.non th s.
The calibration shall be cnecked each effective full power conth usine the IMCORE SYST2:' and recalibrated if the dif ference is > 15 The nin-imum requirements per flu: map used for the recalibration are:
1.
At least 16 different Q
g%
thinble traces, and bE 2.
At least 2 differenct h
thinble traces, per quadrant.
b.
If requirement 3.2.2.C.l.a b.
Not Applicable cyjg cannot be net, then power shall c==::
be linited to 90% of rated crib nower for 4 looo coeration
'and 60% of rate'd h er for h
3 loop operation.
49
),
)
)..
SURVEILLANCE REQUIREMENT LIF.ITING CONDITION FOR OPERATION 3.2.2.C.2.
Inoperable Excore Detector Channel 4.2.2.C.2.
Inoperable Excore Detector Channel If an excore detector channel is If an excore. detector channel l
inoperable, quadrant power tilt
.is inoperable, quadrant power tilt shall be determined by shall be determined by periodically l
monitoring incore thermocouples.
monitoring at least four thermocouples per quadrant once an hour and after any load change greater than 10% at any power level above 50%.
e 3.
NIS Detector Temperature Control 3.
NIS Detector Temperature Control a.
One of the two reactor cavity a.
Reactor cavity ventilation ventilation fans (Unit I fan operation shall be 1RVO12-1A, 1RVOl3-1B) or (Unit II verified once a shift.
2RVO12-2A, 2RVOl3-2B) shall be operating whenever Tavg is great-er than 145'F.
b.
If this condition cannot be met, b.
Not Applicable.
the reactor shall be brought to the Hot Shutdown Condition immediately.
S3
O' O
O-LIMITINd CONDITION FOR OPERATION SURVEILLANCE REOUIRE:*.ENT 302.3.
Control Rod System Operability (per unit) 4.2.3.
Control Rod System Operability (per unit)
A.
Rod Misalignment Limitations A.
Rod Misalignment Limitations 1.
If a full-length control rod 1.
A rod malposition check shall is more than 15 inches (24 steps) be made once a shift using both out of alignment with its bank, the analog and digital displays.
then within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, one of the follcwing steps shall be taken:
a.
Realign the rod, or b.-
Determine by measurement the core peaking factors and apply Specification 3.2.2.A or c.
Restrict power level to 80%
of rated power.
2.
If the misaligned control rod is 2.
Not Applicable not realigned within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, e
the rod shall be declared inoper-able and the limitations of i
3.2.3.B apply, 3.
The provisions of Specifications 3.
Not Applicable j
3.2.3.A and 3.2.3.B shall not i
apply during physics tests in which the control rods are intentionally misaligned.
B.
Inoperable Rod Limitations B.
Inoperabl Rod Limitations s
3.
An inoperable control rod is a 1.
Not Applicable rod which cannot be moved by its mechanism or which is declared inoperable by Specification 3.2.3.A or 3.2.3.C.
51 j
O' O
O
~
LIIIITII;d CONDITION FOR OPERATION SURVEILLA::CE REQUIRE:*ENT 3.2.3.B 2.
Not more than one inoperable 4.2.3.B 2.
Not Applicable.
control rod shall be permitted during power operation.
If more than one rod is determined to be inoperable, the reactor shall be placed in the hot shutdown condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.
If more than one control rod is 3.
Not Applicable.
inoperable because of a Rod Urgent Failure in the rod control system, the provisions of Specifications 3.2.3.B.1 and 3.2.3.B.2 above shall not apply.
If the affected assemblies cannot be returned to service within two hours, the reac-tor shall then be placed in the hot shutdown condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
l 4.
Deleted 4.
Deleted 5.
If an inoperable full-length rod 5.
Not Applicable.
is located above the 200 step level and is capable of being tripped, then the insertion limits in Figure 3.2-2 for Unit I and Figure 3.2-4 for Unit II shall apply for 4 loop operation and the insertion limits in Figure 3.'2-3 for Unit I and Figure 3.2-5 for Unit II shall apply for 3 loop operation.
t 52
The procedures for axial powar distribution conditions for mansuring target flux differenca every control referred to cbove are designed to month.
For this reason, the apscification provides minimize the effects of xenon redistribu-two methods for updating the target flux difference.
tion on the axial power distribution during The alarms provided are derived from the plant pro-loadfollow maneuvers.
Basically control cess computer which determines the one minute averages of flux difference is required to limit of the operable excore detector outputs to monitor the difference between the current value oI in the reactor core and alerts the operator when of Flux Difference ( AI) and a reference AI alarm conditions exist.
Two types of alarm messa-value which corresponds to the full power ges are output.
Above a preset power level, an alarm equilibrium value of Axial offset (Axial message is output immediately upon determining a Offset =d1I/ fractional power;.
The delta flux exceeding a preset band about a target reference value of flux difference varies delta flux value.
Below this preset power level, an with power level and burnup but expressed alarm message is output if the oI exceeded its allow-as axial offset it varies only with burnup, able limits for a preset cumulative amourt of time in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For periods during which the l
The technical specifications on power dis-alarm on flux difference is inoperable, manual sur-tribution control assure that the FQ limit veillance will be utilized to provide adequate warn-is not exceeded and xenon distributions ing of significant variations in expected flux cre not developed which at a later time, differences.
However every attempt should be made would cause greater local power peaking to restore the alarm to an operable condition as soon even though the flux difference is then as possible.
Any deviations from the target band within the limits specified by the proce-during manual logoing shall be treated as deviations dure.
during the entire preceeding logging interval and approprate actions shall be taken.
This action is necessary to satisfy NRC requirements; however more The target (or reference ) value of flux frequent readings may be logged to minimize the difference is determined as follows.
At penalty associated with a deviation from the target cny time that equilibrium xenon conditions band to justify continued operation at the current have been established, the indicated flux power.
l difference is noted with the full length rod control rod bank more than 190 steps The times that deviations from the band occur are withdrawn (i.e. normal full power operating normally accumulated by the computer.
position appropriate for the time in life, usually withdrawn farther as burnup proceeds).
This value, divided by the fraction of full power at which the core was operating,is l
the full power value of the target flux difference.
Values for all other core power levels are obtained by multipl'f ng the full i
power value by the fractional power.
Since
_he indicated equilibrium value was noted, l no allowances for excore detector error are l nncessary and indicated deviation of the A I target band are permitted from the indi-cated reference value.
During periods where extensive load following is required, it may l
be impractical to establish the required core
-68a-
Significantly dif ferent from those resulting from operation within the target brnd.
The instantaneous consecuences of being outside the band, provided rod insertion limits are observed, is not worse tnan a 10 percent increment in peaking fLetor for flux dif ference in the range
-(/+3) percent (+gpercent to -[ percent in-dicated) increasing by +1 percent for each 2 percent decrease in rated powar.
Therefore,
'hile the deviation exists the po.rer level is limited to 90% of PT or lower depcnding on the indicated flux dif ference.
If, for any reason, flux difference is not controlled rithin the AI target bend for as long a period as one hour, then xenon distribution may be significantly changed :nd opcration et 50 percent is re~uired to protect against potentially
- r. ore severe consecuences o-? s ome cccidents.
- s discussed above, the essence of the procedure is to maint=in the xenon distribution in the core a s close to the e~eilibrium full powe.r condition j as possible.
This is accomplished b'; using the boron system to position the fell length control rods to produce the receired indicated flux d if ference.
For Condition II events, the core is protected from overpower and a minimum 2:3n of 1.30 by an automatic protection system.
Compliance with operating procedures is assumed es a pre-condition for Condition II transients; however, operator error and equipment malfunctions are separetely assumed to lend to the cause of transients considered.
- 69a -
In accordance with the approved Westinghouse model DNB Parameters:
The limits on the DNB re-as presented in WCAP 8381, no collapses are expected lated paramters assure that each of the throughout the fuel cycle of operation.
The pre-parameters is maintained within the normal dicted minimum times for clad flattening are:
steady state envelope of operation assumed in the transient and accident analyses.
Fuel Region EFPH The limits are consistent with the initial FSAR assumptions and have been analytically 1
19,500 demonstrated adequate to maintain a minimum 2
> 30,000 DNBR of 1.30 throughout each analyzed 3*
i 30,000 transient.
for Zion Unit 1.
The predicted minimum times to The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these collapse for Unit 2 are:
parameters thru instrument readout is sufficient to ensure that the parameters Fuel Region EPPH are restored within their limits following load changes and other expected transient 1
27,000 operation.
The 18 month periodic measurement of 2
> 30,000 of the RCS total flow rate is adequate to 3*
> 30,000 detect flow degradation and ensure correlation of the flow indication channels with measured A design critariam requires that proposed reload flow such that the indicated percent flow f uel region exposure levels expected at the time will provide sufficient verification of flow of discharge not exceed the predicted minimum rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
collapse time.
Cperation in the exposure range in which clad collapse is postulated is not per-A auadrant ocwer tilt will be ritted under these technical speci fications.
indicated by the excore detectors by the arrangement f the current recorders on the control The predicted ninimum time to collapse for all b
rd.
Four 2-pen recorders are pro-reload fuel regions is greater than 30,000 effee-tive full power hours.
vided, the pens are grouped so that, in the absence of a quadrant power tilt, the two
- Except that the four (4) Region 3 assebmlies to be used in the Extended Burnup progrr. for Zion Unit 2 have a predicted minimum time to collapse greater than 41000 EFpH. -
ink traces coincide.
Any divergence in the traces As described above, an uncertainty fcetor of and indicate a power tilt.
Furthermore, a quadrant 1.08 is included in F H 1.05 in Fg.
There-Ipowertiltalarmisprovidedfortheupperandlow-l fore, a limiting quadrant power tilt of 1.025 er sets of excore currents.
can be tolerated before the margin for uncer-taintyinF8isdepleted, llowever, a measure-Qundrant power tilt limits are based on the follow-ment uncertainty is associated with the indicated ing considerations.
Frequent power tilts are not quadrant power tilt.
Thus, allowing for a low anticipated during normal operation since this l measurement of quadrant power tilt, the action phenomenon is caused by some asymmetric perturba-level of indicated tilt has been set at 1.02, tion, e.g.
rod misalignment, x-y xenon transient, An alarm is set to alert the operator to an or inlet temperature mismatch.
A dropped or l indicated quadrant power tilt of 1.02 or misaligned rod will easily be detected by the Rod greater and that action is required.
To avoid Position Indication System or core instrumentation.
unnecessary power changes, the operator is l A quadrant power tilt by some other means (x-y allowed two hours in which to verify with in-xenon transient, etc.) would not appear instan-core mappinge and/or to determine and correct toneously, but would build up over several hours the cause of the tilt.
Should this action not l cnd the quadrant power tilt limits are set to be taken, the margin for uncertainty in Fd protect against this situation.
They also se.<e is reinstated by reducing the power by 2 per-as a backup protection against the dropped or mis-cent for each percent of tilt above 1.0, in aligned rod.
(8)
Operational experience shows accord with the 2:1 slope envelope described l that normal quadrant power tilts are less than above, or as required by the restriction on 1.01.
Thus, sufficient time is available to rec-peaking factors.
ognize the presence of a tilt and correct the cr.use before a severe tilt could build up.
During l The upper limit on the quadrant power tilt startup and power escalation, however, a large at which hot shutdown is required has been tilt could be initiated.
Therefore, the Technical set at 1.09 so as to provide protection against Specification has been written so as to prevent excessive linear heat generation rate.
escalation above 50 percent power if a large tilt is present.
The numerical limits are set to be commensurate with design and safety limits for DNB protection and linear heat generation rate es described below.
The quadrant power tilt of 1.02 at which remedial cnd corrective action is required has been set to as to provide DNB and linear heat generation rate protection with x-y power tilts.
Analyses htve shown that fractional increases in x-y power peaking factor are less than or equal to twice the increase in the indicated quadrant power tilt, i.e.,
an envelope with a 2:1 slope.
71
In the event that an LVDT is not in service, The nuclear ion chambers located outside a the effects of a malpositioned control rod reactor vessel measure the flux distribution are bservable on nuclear and process infor-of the top and bottom halves of a core.
Core mation displayed in the control room and by traverses in a few of the in-core instrument core thermocouples and in-core movable detec-thimbles will establish that the excore flux tors.
measurement equipment is properly calibrated.
Opsrating experience has established tnat the One inoperable control rod per unit is accep-excore flux measurement system is of a reliable table provided that the power distribution design, and that the 10% load reduction, in the limits are met, trip shutdown capability event of a recalibration delay, is an ultra is available, and provided the potential conservative compensation.
hypothetical ejection of the Inoperable rod is not worse than the case analyzed in the Operating experience at similar PWR plants safety analysis report.
The rod ejection hno shown that quadrant power tilts deter-accident for an isolated fully inserted rod mined by monitoring symmetric thermocouples will be worse if the residence time of the rod are in very good agreement with quadrant is long enough to cause significant non-uni-power tilts determined from power distri-f rm fuel depletion.
The 3 day period allowed bution maps using the Movable Detector f r the analysis is short compared with the System ~
time interval required to achieve a signifi-cant non-uniform fuel depletion.
Operation of one reactor cavity vent fan per unit ensures an adequate flow rate of The rod drop time of 1.8 seconds is based cooling air to each NIS Detector (9).
g
,g g
used in accident analysis.
(11)
The various control rod assemblies (shut-ldownbanks,controlbanksA, B,
C, D) are each to be moved as a bank, that is, with (1)
FSAR - Figure 3.2.1-8 all assemblies in the bank within one step (2)
FSAR - Table 3.2.1-1 (5/8 inch) of the bank position.
Position (3)
FSAR - Figure 3.2.1-11 indication is provided by two methods:
a (4)
PSAR - Chapter 14 digital count of actuation pulses which (5)
FSAR - Section 3.1.2 shows the demand position of the banks and (6)
FSAR - Section 3.1.3 a linear position indicator (LVDT) which (7)
FSAR - Chapter 14, Appendix C indicates the actual rod position (10).
(8)
FSAR - Question 3.8 The rod position indicator channel is (9)
FSAR - Section 9.10.6 sufficiently accurate to detect a mis-(10)
FSAR - Section 7.3 aligned rod 15 inches away from the demand
[8
"]76 der for Modification position of the bank.
This 15-inch (24 steps) permissible misalignment provides f 1,1 cense, an enforceable limit below which design distribution is not exceeded.
72
l l
)
)
)
l LIMITI?:G CONDITION FOR OPERATION SURVEILLA :CE REQUIRE!ET 3.3 3.
Leakage (per unit) 4.3 3.
Leakage (per unit)
A.
If the leakage rate, from other A.
When Reactor Coolant SSstem than controlled leakage sources, pressure is greater than 500 l
such as the Reactor Coolant Pump psig, one of the following Controlled Leakage Seals, exceeds monitoring requirements shall 1 gpm and the source of the leak-be performed (4. 3. 3. A.1 or age is not identified within twen-4.3.3.A.2):
ty-four hours of detection, the reactor shall be brought to hot l
1.
Containme'nt activity shall shutdown within four hours.
If be continuousl/ conitored the source of the leakage is not by radiation detectors identified within an additional 24 RE-00llA or RD 0^12A.
hours, the reactor shall be brought to a cold shutdown condition 2.
Manual sampling of the con-within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
tainment atmosphere shall be performed once a shift.
B.
If the sources of leakage are B.
When Reactor Coolant System identified and the results of the pressure is greater than 500 evaluations are that continued psig at least three of the operation is safe, operation of following monitoring require-the reactor with a total leakage, ments shall be performed (4. 3.3.
other than leakage from controlled B.1, 2,
3, 4, and 5):
sourecs, not exceeding 10 gpm shall be permitted except as specified in 3.3.3.C below.
95
D O
O' LIMITII;d CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT L
3.3.3.B 4.3.3.B 1.
The amount of Reactor Coolant System makeup water required to maintain pressurizer level and volume control tank level shall be recorded.
2.
Containment sump and reactor cavity sump water accumulation shall be monitored daily.
3.
Containment pressure, temperature and humidity shall be monitored.
4.
The high temperature alarm (TE-401) in the reactor head flange leakoff piping shall be operable.
S.
The Reactor Vcssel Leak -
Detection. system (RE-PRl 2A,
RE-PR12B, RY-PRl2A, and associated alarms) shall be operable.
C.
If it is determined that leakage C.
If the monitoring performed in exists through a.non-isolable l
sections 4.3.3.A and 4.3.3.B fault which has developed in a indicates significant leakage Reactor Coolant System component a detailed investigation shall body, pipe wall, vessel wall, or be performed to identify the pipe weld, the reactor shall be sources and quantity of leakage.
brought to a cold shutdown con-dition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and corrective action taken prior to resumption of unit operation.
96 l
l O
O O~
LIMITINd CONDITION FOR OPERATION SURVEILLANCE RECUIREMENT 3.7 STEAM GENERATOR EMERGENCY IIEAT REMOVAL 4.7 STEAM GENERATOR EMERGENCY HEAT REMOVAL Applicability:
Applicability:
l f
Applies to auxiliary feedwater system and Applies to surveillance of auxiliary l
steam generator safety valves, per unit.
feedwater system, and steam generator safety valves per unit.
l Objective:
Objective:
To insure adequate plant cooldown capabil-To insure availability of the above ity upon loss of norinal feedwater flow and system and valves.
loss of main condenser vacuum.
Specification:
Specification:
1.
Steam Line Safety Valves l
1.
Steam Line Safety Valves A.
Twenty ASME code safety valves A.
Ten steam ge'nerator safety valves (5 per steam generator) shall be per unit shall be tested for set operable whenever the reactor is pressure at each refueling' outage.
heated above 350*F except as Testing shall be done by a -
specified in 3.7.1.C, 3.7.1.D, calibrated auxiliary lifting device and 3.7.1.E.
or by bench testing on compressed gas.
At least'two of the valves tested shall be from each orifice size
("Q" or "R").
All valves on a unit shall have been tested at the end of each second refueling outage.
The valves and the corresponding set pressures and orifice sizes are identified in Table 4.7-1.
B.
Deleted B.
Deleted 156
(
)
)
)
SURVEILLANCE REQUIRalENT LIMITING CONDITION FOR OPERATION l
4.8.4.B A flow balance test shall be per-
[
formed on the affected lines during l
the next refueling outage following l
valve stroking or maintenance or other system modifications which might alter the E.C.C.S.
flow characteristics.
E.C.C.S.
flow rates for single pump operation shall meet the following requirements under the minimum resistance configuration:
1.
Charging pum cold leg injection plus seal in. ction shall not exceed 550 GE ;
2.
SI pump hot or cold J injection plus mini flow sL til not exceed 650 GPM; 3.
The minimum charging pump cold leg injection through any 3 l
lines shall be 275 GPM; and 4.
The minimum SI pump cold leg injection through any 3 lines shall be 400 GPM.
173A i
Component Name Component Number Residual Heat Removal Pump-1A (2A)
RH001-1A (2A)
Residual Heat Removal Pump-1B (2B)
RH002-1B (2B)
Residual Heat Exchanger-1A (2A)
RH003-1A (2A)
Residual Heat Exchanger-1B (2B)
RH004-1B (2B)
Recirculation Sump to RHR Pump MOV-SI8811A Suction Valves MOV-SI8811B RWST to RHR Pump Suction Valves MOV-RH8700A MOV-RH8700B Isolation Valves from Reactor MOV-RH8701 Coolant System to RHR Pumps MOV-RH8702 Residual Heat Removal Pumps' MOV-SI8812A Suction Valves MOV-SI8812B Residual Heat Removal Pump System TABLE 4.8-3 187
All fuel rods are pressurized with helium leak rate test.
The structure provides during fabrication to reduce stresses and biological shielding for both normal and strains and to increase fatigue life.
Design Pasis Accident situations.
(1) l Fifty three full length rod cluster control assemblies consisting of 20 individual 80% Ag - 15% In - 5% Cd alloy stainless 5.4.2 Containment System Structure steel clad rods are inserted into the guide thimbles at appropriate locations The Reactor Containment is in the shape of in the core.
a cylinder with a shallow domed roof and a flat foundation slab.
The cylin-Burnable poison rods consisting of drical portion is prestressed by a post-Borosilicate glass sealed in stainless tensioning system consisting of horizontal steel tubes may be used for reactivity and vertical tendons.
The dome has a and/or power distribution control.
three-way post-tensioning system.
The foundation slab is conventionally rein-5.4 Containment System forced with high-strength reinforcing steel.
The entire structure is lined with one-quar-ter inch welded steel plate to provide vapor 5.4.1 Design Basis tightness.
The reactor containment completely The approximate dimensions of the Reactor encloses the entire Reactor Coolant Containment are:
inside diameter, 140 System and assures that essentially feet; inside height, 212 feet; vertical no leakage of radioactive materials to wall thickness, 3-1/2 feet; dome thickness the environment would result even if 2'-8";
and the foundation slab thickness, gross failure of the Reactor Coolant 9 feet.
The containment encloses the System were to occur.
The design of pressurized water reactor, steam generators, the containment liner with channels and reactor coolant loops and portions of the the penetrations permits a much more auxiliary systems and engineered safeguards sensitive and accurate means of
- systems, testing the containment leakage status more frequently than is possible with a conventional integrated
\\
297
_ _ _ = _
6.0 ADMINISTRATIVE CONTROLS 6.1 Organization, Review, Investigation, D.
Qualifications of the station management and Audit and operating staff shall meet minimum acceptable levels as described in ANSI A.
The Station Superintendent shall have N18.1 " Selection and Training of Nuclear overall full-time responsibility for Power Plant Personnel," dated March 8, safe operations of the facility.
1971 with the exception of the Rad-Chem During periods when the Station Supervisor who shall meet or exceed the Superintendent is unavailable, he shall qualifications of Radiation Protection designate this responsibility to an Manager of Regulatory Guide 1.8 established alternate who satisfies September, 1975.
The individual filling the ANSI N18.1 experience requirements the position of Administrative and Support Services Assistant Superintendent shall for plant manager.
meet the minimum acceptable level for B.
The corporate management which relates
" Technical Manager" as described in to the operation of this station is as 4.2.4 of ANSI N 18.1, 1971.
shown in Figure 1-6 of C.E.-1-A Topical Report.
The Division Manager, Nuclear E.
Retraining and replacement training of Stations has responsibility for the Station personnel shall be in accordance Fire Protection Program.
with ANSI N18.1, " Selection and Training of Nuclear Power Plant Personnel," dated C.
The normal functional organization for March 8, 1971.
A training program for the operation of the station shall be as Fire Brigade shall be maintained under the shown in Figure 1-6 and Figure A of direction of the Station Fire Marshal and C.E.-1-A Topical Report.
The shift shall meet or exceed the requirements of manning for the station shall be as Section 27 of the NFPA Code - 1975 except shown in Figure 6.1.1.
The Operating that Fire Brigade training will be con-Assistant Superintendent, Operating ducted quarterly.
Engineer, Shift Engineers, and Shift Foreman shall have a senior operating F.
Retraining shall be conducted at intervals license.
The Fuel Handling Foreman not exceeding two years, has a limited Senior Operating License.
An Operating Engineer will be responsible G.
The Review and Investigative Function and for implementation of the Fire Protection the Audit Function of activities affecting Program.
A Fire Brigade of at least 5 quality during facility operations shall members shall be maintained onsite at all be constituted and have the responsibilities times.
The Fire Brigade shall not include and authorities outlined below:
the minimum shift crew necessary for safe shutdown of the plant (4 members) or any personnel required for other essential functions during a fire emergency.
300
1.
'"he Suocrvisor of the Offsite Review and During the ceriods when the Supervisor cf the Investigative Function shall be appointed by the Offsite Review and Investigative Functier is un-
' lice-President of Construction, Production, available, he shall designate this rescensibility Licensing and Environmental Affairs. The Audit to an establishei alternate who satisfies the formal Function shall be the responsibility of the training and experience requirements fer the supervisor l
Manager of Quality Assurance and shall be of the Offsite Review and Investigative Panction.
independent of operations, a.
Offsite Review a d 'nvestigative Function The responsibilities of the personnel performing this function are stated below. The Offsite Review and The Supervisor of the Offsite Review and Investigative Function shall review:
Investigative function shall:
(i) provide directions for the review and (1) The safety evaluations for 1) chances to procedures, equi rent or systems as described in the safety and investigative function and appoint a 1
senior participant to provide appropriate analysis report and 2) tests or exoeriments direction, (ii) select each participant for corroleted under the provision of 10 CFR Section this function, (iii) select a complement of 50.59 to verify that such actions did not constitute more than one participant who collectively an unreviewed safety question. Proposed changes possess background and qualifications in the to the cuality Assurance Procram description shall subject matter under review to provide be reviewed and approved by the Manager of Ouality
~
comprehensive interdisciplinary review coverage Assurance.
under this function, (iv) independently review rocosed changes to procedures, ecuipment or and approve the findings and recommendations (2) o develooed by personnel performina the review systers which involve an unreviewed safety gaestien and investigative function, (v) approve and as defined in Section 50.59 10 CFR.
report in a tirrely manner all findings of noncompliance with !K recuirerents and provide (3)
Pronosed tests or exoeriments which involve an recomendations to the Station Superintendent, unreviewd safety question as defined in Section Division Manager Nuclear Stations,."anager of 50.59 1C CFR.
Quality Assurance, Vice President of ::uclear Operations, and the Fxecutive '/ ice-President (4) Proposed changes in Technical Specifications or ! E of Construction, Production, Licensinc and operating licenses.
Environmental Affaits.
U U
0 9
EE5 0
9 301 M
Lc57) b
Offsite Review and Investigative Function (Continued)
(5)
Noncompliance with NRC requirements, Such responsibility is delegated to the or of internal procedures or Director of Quality Assurance for instructions having nuclear safety Operating and to the Staff Assistant significance.
to the Manager of Quality Assurance for maintenance quality assurance (6)
Significant operating abnormalities or activities.
deviations from normal and expectea performance of plant equipment that Either shall approve the audit agenda and affect nuclear safety as referred checklists, the findings and the report to it by the Onsite Review and of each audit.
Audits shall be Investigative Function, performed in accordance with the Company Quality Assurance Progrmn and Procedures.
(7)
Reportable Occurrences requiring 24 Audits shall be performed to assure hour notification to the Conmission.
that safety-related functions are covered within a period of two years or
' (- 8 ) - All recognized indications of an as designated below.
unanticipated deficiency in some aspect of design or operation of (1) Audit of the Conformance of facility safety related structures, systems operation to provisions contained or components.
within the Technical Specification and applicable license conditions (9)
Review and report findings and at least once per year.
recommendations regarding all changes to the Generating Stations (2) Audit of the adherence to procedures, Emergency Plan-prior to implemen-training and qualification of the tation of such change.
station staff at least once per year.
(10)
Review and report findings and recommen-(3) Audit of the results of actions dations regarding all items referred by taken to correct deficiencies occuring "3gg the Technical Staff Supervisor, Station in facility equipment, structures, p
3 Suoerintendent, Division Manager -
systems or methods of coeration that Nu' clear Stations and Manager of affect nuclear safety at least once b
3 Quality Assurance.
per six months.
5359 b.
Audit Function (4) Audit of the performance of activities (1
9 required by the Quality Assurance b Eb The Audit Function shall be the responsi-Program to meet the Criteria of bility of the Manager of Quality Assurance Appendix "B",
10 CPR 50.
gg g independent of the Production Department.
(5) Audit of the Facility Emergency Dlan pcad and implementing procedures.
gg 302 b
(6)
Audit of the Pacility Security Plan of Quality Assurance or the Supervisor and implementing procedures.
of the Offsite Reviev and Investigative Function has the authority to order (7)
Audit onsite and offsite reviews.
unit shutdown or request any other action which he deems necessary to avoid (8)
Audit the Facility Fire Protection unsafe plant conditions.
Program and implementing procedures at least once per 24 months.
d.
Records (9)
An independent fire protection and
(.1)
Reviews, audits and recommendations loss prevention program inspection and shall be documented and distributed audit shall be performed at least as covered in 6.1.G.I.a and 6.1.G.1.b.
once per 12 months utilizing either qualified offsite licensee personnel (2)
Copies of documentation, reports, and or an outside fire protection firm.
correspondence shall be kept on file at the station.
(10)
An inspection ~and audit of the fire protection and loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.
(11)
Report all findings of noncompliance with NRC requirements and recommen-dations and results of each audit to the Station Superintendent, the Division Manager-Nuclear Stations, "T50 Manager of Quality Assurance, t' ice President of Suclear Cmrations W
Director of Nuclear Licensing,
(__j and to the Executive " ice President of u
Construction, Production, Licensing, and Environmental Affairs.
O) c.
Authority
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The Manager of Quality Assurance reports b5 c0 l
to the '.' ice Chairman and the sucervisor the Offsite Review and Inve'stigative NEEd Function reports to the Vice President of
))ED Nuclear Operations.
Either the Manager g
302A
e.
Procedures f.
Enrsonnel Written administrative procedures shall be prepared and maintained for the off-site (1)
The persons, including consultante, reviews and investigative functions cerforming the review and investi-described in Specifications 6.1.G.l.a.
gative function, in addition to the These procedures shall cover the following:
Supervisor of the Offsite Review anf Investigative Function, shall have (1)
Content and method of submission of expertise in one or more of the follow-presentations to the Supervisor of the ing disciplines as appropriate for the Offsite Review and Investigative Function, subject or subjects being reviewed and investigated.
(2)
Use of connittees and consultants.
(a) nuclear power plant technology (3)
Review and approval (b) reactor operations
(.c) utility operations (4)
Detailed listing of items to be reviewed.
(d) power plant design (e) reactor engineering (5)
Method of (a) appointing personnel, (f) radiological safety (b) performing reviews, investigations, (g) reactor safety analysis (c) reperting findings and recommendations (h) instrumentation and control of reviews and investigations, (d) approving (i) metallurgy reports, and (e) distributing reports.
(j) any other appropriate disciplines required by unique characteristics (6)
Determining satisf actory completion of of the facility.
action required based on approved findings and reconmendations reported by personnel (2)
Individuals performing the Review and performing the review and investigative Investigative Function shall possess a function.
minimum formal training and experience as listed below for each discipline.
(a)
Nuclear Power Plant Technology Y
Engineering graduate or equiva-g lent with 5 years experience in the nuclear cower field p3-<
design and/or c'peration.
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W 303 E533 h
1
B.
1.
Procedures for items identified in C.
Temporary changes to procedures 6.2.A Specification 6.2.A and any changes and 6.2.B above may be made provided:
to such procedures shall be reviewed l
and approved by the Operating 1.
The intent of the original procedure Engineer and the Technical Staff is not altered.
Supervisor in the areas of operation and fuel handling, and by the Maintenance 2.
The change is approved by two Assistant Superintendent and Technical members of the plant management staff, Staff Supervisor in the areas of plant at least one of whom holds a Senior maintenance, instrument maintenance, Reactor Operator's License on the and plant inspection.
Procedures for unit affected, items identified in Spe cification 6.2.B and any changes to such procedures 3.
The change is documented, reviewed shall be reviewed and approved by the by the Onsite Review and Investigative Technical Staff Supervisor and the function and approved by the Station Rad-Chem Supervisor.
At least one Superintendent within 14 days of person approving each of the above implementation.
procedures shall hold a valid senior operator's license.
In addition, D.
Drills of the emergency procedures these procedures and changes thereto described in Specification 6.2.A.4 shall i
must have authorization by the Station be conducted quarterly.
These drills Superintendent before being implemented.
will be planned so that during the course of the year, communication links 2.
Work and instructions type procedure.n are tested and outside agencies are which implement approved maintenance contacted.
or modification procedures shall be approved and authorized by the Maintenance Assistant Superintendent where the written authority has been provided by the Station Superintendent.
The " Maintenance / Modification Procedure" utilized for safety related work shall be so approved only if "TEd procedures referenced in the" Maintenance /
(__3 i
Modification Procedure" have been g
g aoproved as required by 6.2.A.
5350 Procedures which do not fall within the requirements of 6.2.A or 6.2.B may be approved by the Department Heads.
6 9
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Unit in Operating Mode (other than 3
Position Nono 1 or 2 1& 2 cold shutdown)
Shift Engineer or 1
1 2
Shift Foreman Nuclear Station Operator 1
2 3
Equipment Operator or 2
3 4
Equipment Attendant l
Pad-Crem Technician 1
1 1
TOTAL 5
7 10 MINIMUM
- 5 6
9
- The minimum number refers only for the case of shift shortage, caused by a sudden sickness or home emergency.
Notes:
1.
SRO shall be present on site at all times when there is fuel in the reactor.
2.
A licensed man shall be in the control room at all times whenever fuel is in either reactor.
3.
.Two licensed men shall be in the control room during reactor startups, shutdowns, operation, and other periods such as planned control rod manipulations.
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4.
For the period of Unit 1 and Unit 2 Start up Test b
3 Program, two licensed men per unit shall be in the control room during any operation of the reactor or E35$
l plant which can cause changes in reactivity which have not been verified previously by the startup test program.
(.
3 i
ZION SHIFT MANNING CHART l
Figure 6.1.1 gg g W
331 h
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