ML20003D281
| ML20003D281 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/11/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20003D279 | List: |
| References | |
| NUDOCS 8103260229 | |
| Download: ML20003D281 (29) | |
Text
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e i
c9 fi UNITED STATES 1
NUCLEAR REGULATORY COMMISSION E
,j WASHINGTON, D. C. 20555 D
- a/
THE CONNECTICUT LIGHT AND POWER COMPANY, lHL HAMitVKU LLLLIKlb LibHI LUMPANT, WESILKN MAbbALHU5Lild LLLLIKIL LUMPANT, AND NUKIHLA51 NULLLAN LNLKbT LUMVANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 73 License No. DPR-21 1.
The Nuclear Regulatory Comission (the Comnission) has found that:
A.
The application for amendment by the Connecticut Light and Power Cogany, the Hartford Electric Light Company, Western Massachusetts Electric Comany, and Northeast Nuclear Energy Company (the licensees) dated September 9,1980, as supplemented September.10,1980, November 6, 1980, and February 25, 1980, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission;
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C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health l
and safety of the public; and (ii) that such activities will be conducted in cogliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; i
and l
E.
The issuance of this amendment is in accordance with 10 CFR Part l
51 of the Comission's regulations and all applicable requirements have been satisfied.
l 8103260/N l
t J
l i
l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license aandment, and Paragraph 3.8 of Provisional Operating License No. DPR-21 is hereby amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendix A, as y
revised through Amendment No. 73, are hereby incorporated in the license..The Northeast Nuclear Energy Company shall operate the f acility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION A
prutc) field,LJ Operating Reactors Bran /h'ief' ennis M.
1 ch #5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: March 11, 1981 l
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i L
5 ATTACHMENT TO LICENSE AMENDMENT NO. 73 PROVISIONAL OPERATING LICENSE N0. DPR-21 DOCKET NO. 50-245 Replace the following pages of Appendix A Technical Specifications with the encloseu pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
PAGES,*
3/4 5-7 3/4 6-5 3/4 6-6 3/4 6-10 3/4 7-1 3/4 7-2 3/4 7-15 3/4 11-1 3/4 11-2 3/4 11-3 3/4 11-4 3/4 11-5 3/4 11-6 3/4 11-7 3/4 11-7 a 3/4 11-10 B3/4 4-1 B3/4 5-3 B3/4 6-4 B3/4 7-1 l
B3/4 7-2 B3/4 11-1 Delete 3/4 11-7b l
I i
- 0verleaf pages included for document completeness: 3/4 11-7a, 3/4 11-9, 3/4 7-16, B3/4 4-2, B3/4 6-3
4 4
LIMITING COMITIst FOR OPERMION SURVEILLANCE REQUIREMDIT e
j 2.
From and after the date that one of the 2.
When it is determined that one safety /
l foup reitef/ safety valves of the auto-relief valve of the automatic pressure matic pressure relief subsystem is made or relief subsysten is inoperable the,
found to be Inoperable when the reactor is actuation logic of the remaining APR pressurized above go psig with irradiated valves and FWCI subsysten shall be fuel in'the reactor vessel. reactor opera-demonstrated to be operable tamediately i
tion is permissible only durine the and daily thereafter.
succeeding seven des unless repairs are made and provided that during such time E. Surveillance of the Isolation Camdenser the remaining automatic pressure re11ef Systen shall be perforised as follows:
valves. FWCl subsystem and gas turbine generator are operable.
1.
Isolation Condender Systes Testing:
3.
If the regstraments of 3.5.0 cannot be
- a. The shell side water level and met, an orderly reacter shutdown shall temperature shall be checked be initiated and the reactor shall be daily.
In a cold shutdeun condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l E. Isolation Condenser 51staa 1.
Whenever the reactor pressure is greater than 90 psig and irradiated fuel is in the reactor vessel, the isolation con-i denser shall he operable except as l
spectfled in 3.5.E.2 and the shell side water level shall be greater j
than 66 inches.
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l Amendment No.18 AI, 73 3/4 5-7
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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT coolant system leakage into the primary con-E.
Safety and Relief Valves tainment shall not exceed 25 gpn.
If these
- conditions cannot be met, initiate an orderly 1.
Three of the relief / safety valves top shutdown and have the reactor in the cold works shall be bench checked or replaced shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
with a bench checked top works each re-fueling outage. All six valves top works E.
Safety and Relief Valves shall be checked or replaced every two refueling outages. The set pressure 1.
During power operation and whenever the shall be adjusted to correspond with a reactor coolant pressure is greater than steam set pressure of:
. 90 psig, and temperature greate-than 32@F, the safety valve function of the No. of Valves Set Point (psig) six relief / safety valves shall be operable.
(The solenoid activated relief function of 1
1095 + 11 the relief / safety valves shall be operable 1
1110 T 1%
as required by Specification 3.5.D.)
4 1125 1 1%
2.
If Specification 3.6.E.1 is not met, initiate 2.
At least one of the relief / safety valves an orderly shutdown and have the reactor shall be disassembled and inspected each coolant pressure below 90 psig and tempera-refueling outage.
0 ture below 320 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
During each operating cycle with the reactor 3.
When the safety / relief valves are required at low pressure, each safety valve shall be to be operable per Specification 3.6.E.1, manually opened until operability has been the Acoustic Valve Position Indication shall verified by torus water level instruwenta-be operable. Two of the six channels may tion, or by the Acoustic Valve Position be out of service provided backup indicccion Indication System, or by an audible discharge for the affected valves is provided by the detected by an individual located outside Valve Discharge Temperature Monitor.
the torus in the vicinity of each discharge.
4.
If Specification 3.6.E.3 is cat met, reactor I.
The Acoustic Valve Position Inditation operation is permissible only for the System shall be functionally tested once succeeding 30 days unless the Acoustic every three months, and calibrated once Valve Position. Indication System is made per operating cycle.
operable sooner.
Amendment No. 29, S1, 54, 73 3/4 6-5
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT F.
Structural Integrity F.
Structural Integrity The structural integrity of the primary Inservice Inspection and Testing of primary boundary shall t,e maintained as specified system boundary components shall be performed in Technical Specification 3.13.
as specified in Surveillance Requirement 4.13.
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Amen'dment No. N, 73 3/4 6-6
LIMITING CONDITION FOR DPERATION SINtVEILLANCE REQUIREMENT G.
Jet Pe ps G.
Jet Pumps 1.
Whenever the reactor is in the Startup/ Hot 1.
Whenever there is a recirculation flow Standby or Run modes, all jet peps shall with the reactor in the startup/ hot be intact and all operating jet pumps shall standby or run modes, jet pump inte-be operable.
If it is determined that a grity and operability shall be Checked jet pump is inoperable, an orderly shutdown daily by verifying that the following shall he initiated and the reactor shall be two conditions do not occur in a cold shutdown conditten within 24 simultaneously:
hours.
a.
The recirculation pop flow 2.
Flow indication from each of the twenty jet d6ffers by more than 101 from the pumps shall be verified prior to initiation established speed-flow characteristics.
of reactor startup from a cold shutdown condition.
b.
The indicated total core flow is 3.
the india.ated core flow is the suw of the more than 10% greater than the flow indication from each of the twenty jet core flow value derived from pumps.
If flow indication failure occurs established power-core flow for tw. or more jet pg. hdiate correc-relationships.
tive action shall be taken.
If flow indica-tion cannot be obtained for at least nine-2.
Additionally, wl'en operating with one teen jet pumps, an orderly shutdown shall recirculation pump with the equalizer be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor valves closed, the diffuser to lower shall be in a cold Wutdown con <tition within piene differential pressure shall be checked daily, and the differential
'i 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
pressure of any jet pump in the idle loop shall not vary by more than 101 H.
Recirculation Puesp Flow Mismatch from established patterns.
1.
Whenever both recirculatian pumps are in operation, pump speeds small be maintained 3.
The baseline data required to evaluate within 10% of each other when power level the conditions in Specifications is greater than 80% and within 15% of each 4.6.G.I will be acquired each other when power level is less than 801 operating cycle.
H.
Recirculation pump speed shall be checked daily for mismatch.
AmendmentNo.//,73 3/4 6-10
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LIMITING CIMDITIIM FOR OPERATI0li SURVEILLANCE REQUIREMENT 3.7 Containment Systes 4.7 Contalement Systems I
Applicability Applica bility i
j Applies to the operating status of the primary and AppIles to the primary and secondary coatain-secondary contalment systas.
ment integrity.
I Ob.lective 06.lective l
To assure the integrity of the primary and secondary To verify the integrity of the primary and l
conta1 ament systas.
I f
Spectfication Spect fIcation j
A.
Primary Contalment A.
1.
Suppression Chamber Idater Level and Temperature 1.
The suppression chamber water level and i
bulk temperature shall be checked once l
l 1
The volume and temperature of the water in per shif t.
The interior painted sur-
)
the suppressten chamber shall be maintained faces above the water line of the within the following limits whenever pressure suppression chamber shall j
primary containment is required:
be inspected at each refueling 3
outage.
l a.
Maniaum inter volume 100,400 ft (corresponding to a downcomer a.
Whenever there is indication l
submergency of 3.33 ft. at 1.0 psid) of relief valve operation which j
b.
Minimum inter volume 98,000 ft3 adds heat to the suppression pool.
the bulk pool temperature shall l
(corresponding to a downcomer submer*
be continually monitored and
]-
gence of 3.0 ft. at 1.0 psid) also observed and logged every j
5 minute $ until the heat addition c.
Maximum water taperature:
is terminated.
Amendment No. 13. 51, 73
ListITING CONDITim FOR OPERATION SURVEILLANCE REQUIREMENT (1) During normal power operation -
b.
Whenever there is indication of 90' F.
relief valve operation with the local
. l temperature of the supp ession (2) During testing which adds heat pool reaching 200*F or a. ore l
to the suppression pool, the an external visual examination water teimperature shall not of the suppression chamber shall exceed 10*F above the normal be conducted before resuming i
power operation Ilmit specified power operation.
in(1)above. In connection with such testing, the pool temperature must he reduced to below the normal power operation limit specified in (1) above I
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(3) The reactor shall be scranmed from any operating condition if the pool temperature reaches 110*F. Power operation shall not be resissed untti the pool temperature is reduced $1ow the nonnel power operation Natt spectflod in (1) above.
(4) 'Ouring reactor isolation condi-tions, the reactor pressure vessel shall be depressurized to less than 200 psig at normal cooldoun rates if the pool temperature reaches 120*F.
Amendment No. J3, py, 73 374 72
...-..--[
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TABLE 3.7.1 PRIMARY CONTAlfe1ENT ISOLATION Is31ation Yalve (Valve Number of Power Group Identification Number)
Operated Valves Maximum Action on Operating Initiating Inboard Outboard Time (Sec) Position Signal 1
Main Steam Line Isolation (MS-1A, 2A, 18, 28, IC, 2C, 4
4 3<T<5 0
GC
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10, 2D) l Main Steam Line Drain (MS-5) 1 35 C
SC 1
Main Steam Line Drain (MS-6) 1 35 C
SC 1
Recirculation loop Sample Line (SM-1, 2) 1 1
5 C
SC 1
Isolation Condenser Vent to Main Steam Line IC-6, 7) 2 5
0 GC 2
20 0
GC 2
Dryuell Floor Drain (SS-3, 4) 2 Dryuell Equipment Drain (SS-13,14) 2 20 0
GC 1
10 C
SC 2
Dryuell Vent (AC-7) 1 15 C
SC 2
Dryuell Vent Relief (AC-9) 1 10 C
SC 1
2 Dryuell and Suppression Chamber Vent from Reactor Building (AC-8) 2 Dryuell Vent to Standby Gas Treatment System (AC-10) 1 10 C
SC 2
Suppression Chamber Vent (AC-11) 1 10 C
SC l
2 Seppression Chamber Vent Relief (AC-12) 1 15 C
SC 1
10 C
SC 2
Suppression Chamber Supply (AC-6) 1 10 C
SC 2
Dryuell Supply (AC-5) 2 Dryuell and Suppression Chamber Supply (AC-4) 1 10 C
SC 3
Cleanup Domineralizer System (CU-2) 1 18 0
6C 2
18 0
GC l
3 Cleanup Domineralizer System (CU-3, 28) 3 Shutdoun Cooling System 50-1) 1 48 C
SC 3
Shutdown Cooling System SD-2A,28,4A,48) 4 48 C
SC 3
Shutdoun Cooling System 50-5) 1 48 C
SC 3
Reactor Head Cooling Line (HS-4) 1 45 C
SC 4
Isolation Condenser Steam Supply (IC-1) 1 24 0
GC 4
Isolation Condenser Steam Supply (IC-2) 1 24 0
GC 4
Isolation Condenser Condensate Return (IC-3) 1 19 C
SC i
4 Isolation Condenser Condensate Return (IC-4) 1 19 0
GC Feedster Check Valves (FW-9A,10A, 98,108) 2 2
NA 0
Process Control Rod Hydraulic Return Check Valves (301-95, 98) 1 1
MA 0
rrocess Reactor Head Cooling Check Valves (HS-5) i NA C
Process Stan6y Liquid Control Check Valves (SL-7, 8) 1 1
NA C
Process 3
Cleanup Denineralizer System (CU-5) 1 18 C
SC l
-5 Amendment No. 73
TABLE 3.7.1 Key: 0 = Open C = Closed SC = Stays Closed GC = Goes Closed NA = Not Applicabit Note:
Isolation groupings are as follows:
GROUP 1: The valves in Group 1 are closed upon any one of the following conditions:
1.
Reactor low-low water level (This signal also trips the reactor recirculation pumps.)
2.
Main steam line high radiation.
3.
Main steam line high flow.
4.
Main steam line tunnel high tenperature.
5.
Main steam line low pressure.
GROUP 2: The actions in Group 2 are initiated by any one of the following conditions:
1.
Reactor low water level.
2.
High drywell pressure.
GROUP 3: Reactor low water level alone initiates the following:
1.
Cleanup demineralizer system isolation.
2.
Shutdown cooling system isolation.
3.
Reactor head cooling isolation.
GROUP 4:
Isolation valves associated with the isolation condenser are closed upon indication of either high isolation condenser steam or condensate flow.
3/4 7-16
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11 REACTOR FUEL ASSEMBLY 4.11 REACTOR FUEL ASSEMBLY Applicability Applicability The Limiting Conditions for Operation associated with The Surveillance Requirements apply to the para-the fuel rods apply to those parameters which monitor meters which monitor the fuel rod operation the fuel rod operating conditions.
conditions.
Objective Objective The Objective of the Limiting Conditions for Opera-The Objective of Surveillance Requirements is tion is to assure the performance of the fuel rods.
to specify the type and frequency of surveillance to be applied to the fuel rods.
Specifications Specifications A.
Average Planar Linear Heat Generation Rate (APLHGR)
A.
Average Planar Linear Heat Generation Rate (APLHGR) 1.
During power operation, the APLHGR for each type of fuel as a function of average The APLHGR for each type of fuel as a planar exposure shall not exceed the limit-function of average planar exposure shall ing value shown in Figure 3.11.1.
be determined daily during reactor operation at > 25% rated thermal power.
2.
If at any time during operation it is determined by normal surveillance that the limiting value for APLHGR specified in Section 3.11.A.1 is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown conditiori within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
3/4 11-1 Amendment No. 4, AA, 28,14, 49, 67. 73
3-a s
b 12 5
11.5 11.5 10.9
_ 11.2 U
N
[
10.9
,10.7 10.
5 m
10.7 10.7 10.5 10.6 10.4 10.2 10 z
0.95 multi plice appli ad when oper ating at 9.7 (
< 90% of r ated core f low.
9,3 {
9.1 9
n.
m 8.6 b
5 I
8 5
32,000 36,000 g
0 5000 10,000 15,000 20,000 25,000 30,000 PLANAR AVERAGE EXPOSURE mwd /t Figure 3.ll.la - M/21 MUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS PLANAR AVERAGE EXPOSURE. FUEL TYPE 808262.
Amendment No. 4, M M, 0, 73
t l
^
C s
12 m
g 11.3 11.5 11.2 5
10.9 l
u 10.9 "10.8 U
g 10.7 h
10.4
- 10. 4 10 l
9.9 (
0.95 mult iplier appl-ed when ope ating at zs 9.5 (
< 90% of ated core 1'l ow.
9.2 m
8 w
w 32,000 ' 36,000 0
5000 10,000 15,000 20,000 25,000 30,000 PLANAR AVERAGE EXPOSURE mwd /t Figure 3.11.lb - MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS PLANAR AVERAGE EXPOSURE.
FUEL TYPE 80274L.
Amendment No. X,J4 73 3/4 11-3
Ct 12 w
b "11 5
- 4
_11.3 11.2 8
11.0 10.9 "N
h 11 C
C Nn10.9 g
m m
10.9 o
10.7 10*8 10.7 9
10.5 p
10.6 10.4 10.4 10.0 10 b
9.9 (
0.95 multiplier appli ed when operating at a
9.5 (
< 90% of rated core flow.
9.2 3
9 n
U 8.8
- l!
ra 5
8 a
i 4
32,000" 36, 000 0
5000 10,000 15,000 20,000 25,000 30,000 g
PLANAR AVERAGE EXPOSURE mwd /t Figure 3.11.lc - MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS PLANAR AVERAGE EXPOSURE. FUEL TYPE 80274H.
Amendment No. ff ff, 73
s 12 U
- li II I II I 11.0 11.0 3
11 10.
10 6 g
10.7 u
g 10.5
'10.5 "10.5 10.5 0.2 p
10.2 to jii_7 h
10.1
/
z 0.95 mult iplier app 1' ed when ope ating at 9.7 (
a 9.7 g
< 90% of rated core 1'10w.
9.3- (
s.
(e2 n
g 8.8 N,
a 8
g 32,000 36,000 g
0 5000 10,000 15,000 20,000 25,000 30,000 PLANAR AVERAGE EXPOSURE mwd /t Figure 3.11.1d - MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS PLANAR AVERAGE EXPOSURE. FUEL TYPE 8DR265L.
Amendment No. M 73 3/4 11-5
k1$
12 N
2 k
11.1 11.1 10.9 #)
C y
n 5
10.7
(
'UM
" iv.y E
10.4 10.5 10.5 o
10.2 10.4 "10.4 10 1
,10.2 8
/
C 0.95 multi plier applies when operating at
< 90% of rated core f low.
(
~
9.2-n.
g 8
0 32,000 36,000 0
5000 10,000 15,000 20,000 25,000 30,000 PLANAR AVERAGE EXPOSURE mwd /t Figure 3.ll.le: - MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS PLANAR AVERAGE EXPOSURE. FUEL TYPE 80R265H.
Amendment No. J6, 3#, #9, 67, 73 3/4 11-6 l
J
12 m
5 ll.3 a
. 11.3 11.1 (112 11'2 g
10.8 g' g
j f
7 10.7 J 0.7 10.6 l0.7 jo,3 10 6 m
10.3 eq o
p10.2 10 8
O.95 mul ;iplier appl ied when operating at 9.7 g
< 90% of rated core flow.
9.3 (
9.2 B.8 m
a*
k g
,x.
u.
0 5000 10,000 15,000 20,000 25,000 30,000 PLANAR AVERAGE EXPOSURE mwd /t Figure 3.11.lf - MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR)
VERSUS PLANAR AVERAGE EXPOSURE.
FUEL TYPE P80R265H.
4 Amendment No. 34 pg, 73
n C
s 12 l
U 11.3 all.3 11.2
[11j "
m r,
O h
10.7 j b
10.7 10.7 11.0 y 0.6 10.2 d10 5 10.2 g
10 jg_n g
0.95 multi > lier applie d when opertting 3
9.Q a
94 at < 90%4)f rated core flow.
ac on 9
U
- i s
8 32,000 36.000 0
5000 10,000 15,000 20,000 25,000 30,000 g
PLANAR AVERAGE EXPOSURE mwd /t Figure 3.11.1g - MAXIMUM AVERAGE PLAMAR LINEAR HEAT GENERATION RATE (MAPuiGE) i VERSUS FIAm AVERACF EXPOSURE.
FUEL TYPE P80R282.
Amendment no. (1. g, 73 3/4 Il-7a
LislTING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Minlede Critical Power Ratio (MCPR)
C.,
MinimumCriticalPowerRatio(MCPR1 C
During power operation MCPR shall be as shown in hCPR shall be determined daily during Tal.le 3.11.1 If at any time during operation reactor power operation at > 25% rated it is determined by normal surveillance that the thermal power and following any change in 1
limiting value for MCPR is being exceeded, action power level or distribution that would cause l
shall be initiated within 15 minutes to restore operation with a limiting control rod pattern I
operation to within the prescribed limits. If as described in the bases for specification the steady state MCPR is not returned to within
'3.3.B.5.
the prescribed limits within two (?) hours, the reactor shall be brought to the t:old Shutdown condition witliin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue un(11 reactor operation is within the prescribed limits.
For core flows other than rated time MCPR's in Table 3.11.1 shall be multiplied by K, where g
K is as shown in figure 3.11.2.
g D._
If any of the limiting values identified in Specificattuns 3.ll.A. B, or C. are exceeded, even if corrective action is taken, as pre-scribed, a Reportable Occurrence report shall be submitted.
i 1
- u. ia.nt uu. #, Jp, en 3/4 11-9 l
TABLE 3.11.1
(
OPERATING LIMIT MCPR'S FOR CYCLE 8 BOC8 TO 6000 mwd /t 6000 mwd /t TO EOC8 EOC8 TO 70% C0ASTDOWN FUEL TYPE 1.31 1.37 1.37 8x8 1.29 1.37 '
l.37 8 x 8R 1.31 1.39 1.39 P8 x 8R f
0 t
[
Amendment No. /s, g, //, #, 73 3/4 11-10
3.4 Bases
The design objective of the liquid control system is to provide the capability of brini y the reactor from A.
To full power to a cold menon-free shutdown assuming that none of the withdrawn control rods can be inserted.
meet this objective, the lipid control system is designed to inject a quantity of boron which produces a concentration of.660 ppu of boron in the reactor core in less than 125 minutes. The 660 ppm concentration in the reactor core'would bring the reactor from full power to a minimum 2.6% delta K subcritical condition an:) uncertainties, etc.
considering the hot and cold reactivity swing. xenon poisonf rvj analytical biases An additicr.a125% of borem solution is provided for possible imperfect mixing of the chemical solution in the reactor coolant. A minimum quantity of 2720 net gallons of solution having a 13.4% sodim pentaborate concentra-Actual systm volume for this quantity is 2960 gallons.
tjon is required to meet this shutdown requirement.
(240 gallons are contained below the pump suction and, therefore, cannot be inserted.)
The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of For the minimum required reactivity insertion due to cooldown of the reactor following the menon poison peak.
pumping rate of 32 gallons per minute, the maxisum storage volume of the boron solution is established as l
4190 gallons.
Boron concentration, salution temperature. (within the tank and connecting piping including check of tank heater I
4pd pipe heat tracing systa) and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Experience with pimp operability indicates that monthly testing ts adequate to detect if failures have occurred.
Components of the syste are checked periodically as described above and make a functional test of the entire A test of one installed system on a frequency of less than once during each operating cycle unnecessary.
explosive charge is made at least once during each operating cycle to assure that the charges are ratisfactory.
A continual check of the Tha replacement charge will be selected from the same batch as the tested charge.
l firing circuit continuity is provided by pilot lights in the control room.
The relief valves in the standby liquid control system protect the system piping and positive displacemnt punps which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank.
If l
B.
Ohly one of the two staney itquid control pumping circuits is needed for proper operation of the system.
one puePing circuit is found to be inoperable, there is no imediate threat to shutdown capability, and reactor Operation may continue while repairs are being made. Assurance that the remaining system will perform its l
intended function and that the reliability of the system is good is obtained by demonstrating operation of the 1
pump in the operable circuit at least once daily.
]
Amendment No. 73 B 3/4 4-1 i
i
C.
The solution saturation temperature of 13.4% sodium pentaborate, by weight, is 59*F.
The solution shall be kept at least 10*F above the saturation temperature within the tank and suction piping to guard against boron precipitation. The 10*F margin is included in Figure 3.4.2.
Temperature and liquid level alarms for the
. system are annunciated in the control room.
Pump operability is checked on a frequency to assure a high reliability of operation of the system should it cver be required.
Once the' solution has been made up, boron concentration will not vary unless more boron or more water is added
- cr removed.
Level indication and alarm indicate whether the solution volume has changed which might indicate a possible solution concentration change. Considering these factors, the test interval has been established.
l j
e 0
8 3/4 4-2
without the use of off-site electrical power.
For the pipe breaks for which the FWCI is intended to function, the core never uncovers and is continuously cooled and thus no clad damage occurs. The repair times for the limiting conditions of operation were set considering the use of the FWCI as part of the emergency core cooling system and isolation cooling system.
The FWCI utilizes portions of the normally operating feedwater system; e.g., condensate, condensate booster and fecdwater pumps. Therefore, the reliability of the pumps. valves and motors is constantly being demonstrated.
Thus the system has an inherently higher degree of reliability than normally non-operating systems. Since an operating string of pump and valves is programed for FWCI operation, it is not expected that the normally operating portions of the FWCI would be out of operation during normal operation. Thus, to demonstrate the operability of the FWCI, it is usually sufficient to demonstrate the non-operating portions of the system.
D.
Automatic Pressure Relief (APR)
The relief valves of the automatic pressure relief subsystem are a back-up to the FWCI subsystem. They enable the core spray or LPCI to provide protection against the small pipe break in the event of FWCI failure, by depressurizing the reactor vessel rapidly enough to actuate the core sprays or LPCI. The core spray and/or LPCI provide sufficient flow of coolant to limit fuel clad temperatures to well below clad melt and to assure
~
that core geometry remains intact.
APR testing at low reactor pressure is required during each operating cycle.
It has been demonstrated that t,.a tiowdown of the APR to the torus causes a wave action that is detectable on the torus water level instru-menuation. The discharge of a relief line is audible to an individual located outside the torus in the vicinity of the line, as experienced at other BWR's.
E.
Isolation Condenser System The turbine main condenser is normally available. The isolation condenser is provided for core decay heat removal following reactor isolation and scram.' The isolation condenser has a heat removal capacity sufficient to handle the decay heat production at 300 seconds following a scram. Water will be lost from the reactor vessel through the relief valves in the 300 seconds following isolation and scram. This represents a minor loss relative to the vessel inventory.
Amendment No. 4, 73 B 3/4 5-3
_~
HoNever,Lthere are various conditions under which the dissolved oxygen content of the reactor coolant water could he higher than 0.2-0.3 ppm, such as refueling and reactor startup.
During these periods with steaming rates less_than.100,000 pounds, per hour,'a-more restrictive limit of 0.1 ppm has been established to assure ths chloride-oxygen combinations are maintained at conservative levels. At steaming rates of at least 100,000 pounds per-hour, boiling occurs causing deaeration of the reactor water, thus maintaining oxygen
.c:ncentration at low levels.
Wh n conductivity is in its proper normal range, pH,and chloride and other impurities affecting conductivity must also be within their normal range. When conductivity becomes abnormal, then chloride measurements are
.cade to determine whether_ or not they are also out of their normal operating values.
This would not n:cessarily be the case. Conductivity could be high due to the presence of a neutral salt; e.g., Na250,
4 which would not have an effect on pH or chloride.
In such a case, high conductivity alone is not a cause for shutdown.
In some types of water-cooled reactors. conductivities are in fact high due to purposeful addition of additives.
In the case of BWRs, however, where no additives are used and where neutral pH is maintained, i
ccnductivity provides a very good measure of the quality of the reactor water.
Significant changes therein provide the operator with a warning mechanism so he can investigate and renedy the conditien causing the change.before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceeded. Methods available to the operator for correcting the off-standard condition include operation of the reactor cleanup system, reducing the input of impurities and placing the reactor in the cold shutdown ccndition. The major benefit of cold shutdown is to reduce the temperature dependent corrosinn rates and provide time for the cleanup system to re-establish the purity of the reactor coolant.
During startup periods, which are in the category of less -than 100,000 pounds per hour, conductivity may exceed 2 pmho/cm because of the initial evolution of gases and the initial addition of dissolved metals.
During this period of time, when the conductivity exceeds 2 umho (other than short-term spikes), samples will be taken to assure that the chloride concentration is less than 0.1 ppm.
The conductivity at the reactor coolant is continuously monitored.
The samples of the coolant which are taken every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors.
If conductivity is within its normal range, chlorides and other impurities will also be within their normal ranges. The reactor coolant samples will also be used to deter-cine the chlorides. Therefore, the sampling frequency is considered adequate to detect long-tenn changes in the* chloride ion content.
Isotopic, analyses to' determine major contributors to activity can be performed by 2
a gamma scan.
L D.
Ccolant Leakage The 0.5 gpm limit for leaks from unidentified sources was established by assuming the leakage was from the primary system. Tests demonstrate that a relationship exists between the size of a crack and the probability 1
that a crack will propagate.
1 8 3/4 6-3
Frr a crack siz2 which givss a leakags rate of 2.5 gpm, tha probability of rapid propagation is less than 10-5 A leakage rate of 2.5 gpm is detectable and measurable.
The 25 gpm limit on total leakage to the containment was established by considering the removal capabilities of the pumps.- The capacity of either of the drywell sump pumps is 50 gpm and the capacity of either of the-drywell equipment drain tank pumps is also 50 gpm. Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
The performance of the reactor coolant leak detection system will be evaluated during the first year of commercial operation and the conclusions of this evaluation will be reported to the AEC.
The main steam line tunnel leakage detection system is capable of detecting small leaks. The system per-formance will be evaluated during the first five years of plant operation and the conclusions of the evalu-ation will be reported to the AEC.
E.
Safety and Relief Valves Present experience with the new safety / relief valves indicates that testing of at least 50% of the safety valves per refueling outage is adequate to detect failures or deterioration. The tolerance value is speci-fled in Section III of the ASME Boiler and Pressure Vessel Code as +1% of design pressure. An analysis has be(n performed which shows that with all safety valves set 1% higher the reactor coolant pressure safety liutt of 1375 psig is not exceeded.
The relief / safety valves have two functions:
1.e., power relief or self-actuated by high pressure.
The solenoid actuated function (automatic pressure relief) in which external instrumentation signals of coinci-dent high drywell pressure and low-low water level initiate the valves to open.
This function is discussed in Specification 3.5.D.
In addition, the valves can be operated manually.
The safety function is performed by the same relief / safety valve with a pilot valve causing main valve operation.
It is understood that portions of the Acoustic Valve Position Indication cannot be repaired or replaced 4
during operation, therefore, the plant must be shutdown to accomplish such repairs.
The 30-day period
- to do this allows the operator the flexibility to choose his time for shutdown; meanwhile, because of the redundancy provided by the valve discharge temperature monit.or and the continued monitoring of the remaining valves by both methods, the ability to detect the opening of a safety / relief valve would not be compromised. The valve operability is not affected by failure of the Acoustic Valve Position Indi-cation System.
Amendment No. AT, 73 B 3/4 6-4
A.
1.
Primary Containment The integrity of the primary containment and operation of the emergency core cooling system in combination, limit the off-site doses to values less than those specified in 10 CFR 100 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the pctential for violation of the primary reactor system integrity exists.
Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure.
An exception is made to this requfrement during initial core loading and while the low power test program is being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at this time which will greatly reduce the chances of a pipe break. The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring.
Procedures and the Rod Worth Minimizer would limit control worth to less than 1.5% AK. A drop of a 1.5%oK rod does not result in any fuel damage.
In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well within 10 CFR 100 guideline values.
2.
Suppression Chamber The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system or for releases through the safety relief valves. The pressure l
suppression chamber water volume must absorb'the associated decay and structural sensible heat released during primary system blowdown from 1035 psig.
Since all of the gases in the drywell are considered purged into the pressure suppression chamber air space
}
during a loss of coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber design pressure.
The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in the specification, containment pressure during the design Maximum water volume of basisaccigentisapproximately42psigwhichisbelowthedesignof62psig.
3 results 100,400 ft results in a downcomer submergence of 3.33 feet and, the minimum volume 98,000 ft in a submergence of 3.0 feet. The majority of the Bodega tests were run with a submerged length of four feet and with complete condensation. Additional condensation tests were run in the Mark I Full Scale Test Facility (FSTF) at downcomer submergence varying between 1.5 and 4.5 feet and complete condensation of steam resulted. Thus, with respect to downcomer submergence, this specification is adequate.
The maintenance of a drywell-suppression chamber differential pressure of 1.00 psid and a suppression chamber water level corresponding to a downcomer submergence range of 3.0 to 3.33 feet will assure the Post-LOCA suppression pool swell hydrodynamic forces are minimized and consistent with loads assumed for structural analysis of the suppression chamber.
I AmendmentNo.ff,73 8 3/4 7-1
III and Bodega Bay (2) tests was The maximum temperature at the end of blowdown tested during the Humboldt Bay
~
170*F and this is conservatively taken to be the limit for complete condensation of the reactor coolant.
L Tests done in the FSTF showed complete condensation with bulk temperature as high as 185*F with a corresponding surface temperature of 230'F.
Regarding condensation of steam released through the SRVs and quenchers, test data has shown complete condensation beyond the 200'F limit of the NRC Acceptance Criteria.
3 i
Based on the minimum water volume of 98,000 ft that the Millstone suppression pool contains, the expected bulk pool temperature rise during the reactor blowdown is less than 70'F.
With an initial pool temperature of 90"F, considerable margin exists between this postulated pool temperature and the temperature for which complete condensation was demonstrated.
For an initial maximum suppression chamber water temperature of 90'F and assuming the normal complement of pumps (2 LPCI pumps and 2 emergency service water pumps) in each loop, containment pressure is required during a small period of the total accident to maintain adequate net positive suction head (NPSH) for the core spray and LPCI The availability of the containment pressure required to maintain adequate NPSH during this interval is pumps.
assured by the containment spray interlocks as described in Amendment 18.
If a loss of coolant accident were to occur when the reactor water temperature is below 330*F, the containment The maximum pressure will not exceed the 62 psig design pressure, even if no condensation were to occur.
allowable pool temperature, whenever the reactor is above 212*F shall be governed by this s)ecification. Thus, specifying water volume tenperature requirements applicable for reactor water temperatures a)ove 212*F provides additional margin above that available at 330'F.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define 1
This action would the action to be taken in the event'a relief valve inadvertently opens or sticks open.
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat include:
exchangers, and (3) initiate reactor shutdown.
Robbins, C. H., Tests of a Full Scale 1/48 Segment of the Humboldt Bay Pressure Suppression Containment, GEAP-3596, November 16, 1960.
Bodega Bay Preliminary Hazards Sunsnary Report, Appendix 1, Docket 50-205, December 28, 1962.
l NEDE 24539P Mark I Containment Program Full Scale Test Program Final Test Report, April 1979.
NED0 24575, Mark I Containment Program Plant Unique Load Defnition, Millstone Nuclear Power Station -
Unit 1, March 1979.
NRC Acceptance Criteria for Mark I Containment Long Term Program, Rev.1, Februa.y 1980.
NUREG-0661 Mark I Containment Long Term Program, July 1980.
Amendment No. U, 57, 73 8 3/4 7-2
3.11 and 4.11 Bases A.
Average Planar Linear Heat Generation Rate (APLHGR)
This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10 CFR 50, Appendix K.
The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than + 20*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rato is sufficient to assure that. calculated temperatures are within the 10 CFR 50 Appendix K limit. The limiting value for APLHGR is shown in Figure 3.11.1.
Conservative LOCA calculations predict that nucleate boiling will be maintained for several seconds following a design basis LOCA. This results in early removal of significant amounts of stored energy which, if present later in the transient, when heat transfer coefficients are cor.siderably lower, would result in higher peak cladding temperature. As core flow is reduced below about 90%, the time of onset of boiling transition makes a sudden change from greater than about 5 seconds to less than i second. The approved ECCS evaluation model requires that at the first onset of local boiling transition, the severely reduced heat transfer coefficients must be applied to the affected planar area of the bundle, and thus exaggerates the calculated peak clad temperature. The effect is to significantly reduce the energy calculated to be removed from the fuel during blowdown. This results in an increase in calculated peak clad temperature of about 100*F which can be offset by a 5% reduction in MAPLHGR.
For flows less than 90% of rated, a 5% reduction in the MAPLHGR limits in Figure 3.11.1, derived for 100% flow will assure that the plant is operated in compliance to 10 CFR 50.46 at those lower flows.
B.
Linear Heat Generation Rate (LHGR)
This specification assures that the linear heat generation rate in any rod is less than the design linear heat generation rate if fuel pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of References 1 and in Reference 2, 3, and 4 and assumes a linearly increas-ing variation in axial gaps betweer core bottom and top, and assures with a 95% confidence, that no more than one fuel rod exceeds the desijn linear heat generation rate due to power spiking. The LHGR as a function of core height shall be checked daily during ' reactor operation at > 25% power to determine if fuel Amendment No. J, 75, 79, )(, M, 57, 73 8 3/4 11-1
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