ML20211F199
| ML20211F199 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/22/1986 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211F203 | List: |
| References | |
| NUDOCS 8610310098 | |
| Download: ML20211F199 (18) | |
Text
!aun 'o UNITED STATES 8",
/,'n NUCLEAR REGULATORY COMMISSION
{
- ,E WASHINGTON, D. C. 20555
\\...../
NORTHERN STATES POWER COMPANY DOCKET N0. 50-263 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 47 License No. DPR-22 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Northern States Power Company (the licensee) dated March 24 and July 22, 1986 complies with the standards and requirements 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 application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance 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; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows:
8610310098 861022 PDR ADOCK 05000263 P
- * ^ -
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. (2) Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 47, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULAT RY COMMISSION t
l s.
John. Zwolinski, Director BWR P o,iect Directorate #1 Divisi n of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 22, 1986 l
.....n..
ATTACHMENT TO LICENSE AMENDMENT NO. 47 FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 Revise Appendix "A" Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain marainal lines indicating the area of change.
REMOVE INSERT 11 11 v
v 6
6 20 20 56 56 58 58 113 113 114 114 114a 114b 119 119 211 211 213 213 214 214 218 218
Page 3.4 and 4.4 Standby Liquid Control System 93 A.
Normal Operation 93 B.
Operation with Inoperable Components 94 C.
Volume-Concentration Requirements 95 3.4 and 4.4 Bases 99 3.5 and 4.5 Core and Containment Cooling Systems 101 A.
Core Spray System 101 3.
LPCI Subsystem 103 C.
RHR Service Water System 106 D.
HPCl System 108 E.
Automatic Pressure Relief System 109 F.
RCIC System 111 G.
Minimum Core and Containment Cooling System Availability 112 H.
Recirculation System 114 1.
Deleted 3.5 Bases 115 4.5 Bases 120 3.6 and 4.6 Primary System Boundary 121 A.
Reactor Coolant Heatup and Cooldoun 121 B.
Reactor Vessel Temperature and Pressure 122 C.
Coolant Chemistry 123 D.
Coolant Leakage 126 l
E.
Safety / Relief Valves 127 F.
Deleted i
G.
Jet Pumps 128 i
H.
Snubbers 129 3.6 and 4.6 Bases 144 3.7 and 4.7 Containment Systems 156 A.
Primary Containment 156 B.
Standby Gas Treatment System 166 C.
Secondary Containment 169 D.
Primary Containment Isolation Valves 170 E.
Combustible Gas Control System 171a 3.7 Bases 175 4.7 Bases 183 11 Amendment No. 9,33, 47'
LIST OF FIGURES Figure No.
Page No.
4.1.1 "M" Factor - Graphical Aid in the Selection of 44 an Adequate Interval Between Tests 4.2.1 System Unavailability 75 3.4.1 Sodium Pentaborate Solution Volume-Concentration 97 Requirements 3.4.2 Sodium Pentaborate Solution Temperature Requirements 98 3.5.1 Single loop Operation Surveillances Power / Flow Curve 114b 3.6.1 Chanae in Charpy V Transition Temperature versus 133 Neutron Exposure 3.6.2 Minimum Temperature versus Pressure for Pressure 134 Tests 3.6.3 Minimum Temperature versus Pressure for Mechanical 135 Heatup or Cooldown Following Nuclear Shutdown 3.6.4 Minimum Temperature versus Pressure for Core 136 Operation 4.6.2 Chloride Stress Corrosion Test Results 0 500 F 137 3.7.1 Differential Pressure Decay Between the Drywell 191 and Wetwell 3.8.1 Monticello Nuclear Generating Plant Site Boundary 198o for liquid Effluents 3.8.2 Monticello Nuclear Generating Plant Site Boundary 198h for Gaseous Effluents 3.11.1 MAPFAC Limits 215a p
3.11.2 MAPFAC Limits 215b p
3.11.3 Power Dependent MCPR Limits 215c 3.11.4 MCPR Limits 215d p
6.1.1 NSP Corporate Organizational Relationship to On-Site 234 Operating Organization 6.1.2 Monticello Nuclear Generating Plant Functional 235 Organization for On-Site Operating Group Amendment No. 29, A(-' 47 v
2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1 FUEL CLADDING INTEGRITY 2.3 FUEL CLADDING INTEGRITY Applicability Applicability Applies to the interrelated variables Applies to trip settings of the instruments associated with fuel thermal behavior, and devices whlch are provided to prevent the reactor system safety limits from being exceeded Objective:
Objective:
To establish limits below which the integrity To define the level of the process variables of the fuel cladding is preserved.
at which automatic protective action is initiated to prevent the safety limits from being exceeded.
Specification:
Specification:
A.
Core Thermal Power Limit (Reactor The Limiting safety system settinga shall be Pressure >800 psia and Core Flow is >10 %
as specified below:
of Rated)
A.
Neutron Flux Scram When the reactor pressure is >800 psia and core flow is >10 t of rated, tius
ratio (MCPR) less than 1.07, for two recirculation loop operation, or less than S < 0.58(W-dw) + 62%
1.08 for single loop operation shall constitute violation of the fue,l cladding
)
- where, itlegrity safety limit.
S - Setting in percent of rated thermal power, rated power being 1670 MWT jf W - Recirculation drive flow in g
percent Et dw - O for two recirculation loop g
operation
- 5.4 for one recirculation loop E.i operation.
f3 2.1/2.3 6
Bases Continued:
that the reactor mode switch be in the startup position where protection of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scram. Thus the combination of main steam line low pressure isolation and isolation valve closure scram assures,the availability of the neutron scram protection over the entire range of applicability of the fuel cladding integrity safety limit.
The operator will set this pressure trip at greater than or equal to 825 psig. However trip setting can be as much as 10 psi lower due to the deviations discussed on page 39., the actual References
- 1. Linford, R. B., " Analytical Met. hods of Plant Transient Evaluations for the General Electric Boiling Water Reactor". NEDO-10802, Feb., 1973.
- 2. " Average Power Range Monitor, Rod Block Monitor and Technical Specifications Improvement (ARTS) Program for Monticello Nuclear Generating Plant", NEDC-30492-P, April,'1984.
- 3. "Monticello Nuclear Generating Plant Single Loop Operation", NEDO-24271, June, 1980.
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2.3 BASES 20
1 TABLE 3.2.3 Instrunentation " nut Initiates Rod Block Reactor Modes in W ich Rrction !bst be Operable Total No. of Min. No of Oper-or Operating arul Allow-Instnanent able or Operatirg able Bypass Conditions **
domels per Instnanent Qunnels Requiral Rartion Trip Settings Refuel Startip Rm Trip Systern per Trip Systan Conditions *
- 1. E 5
a.
Upscale
$ 5x10 cps X
X(d) 2 1 (Note 1, 3, 6)
A or B or C b.
Detector X(a)
X(a) 2 1 (Note 1, 3, 6)
A or B or C tot fully inserted
- 2. E a.
Downscale
> 3/125 X(b)
X(b) 4 2 (Note 1, 4, 6)
A or B or C full scale b.
Upscale
$ 108/125 X
X 4
2 (Note 1, 4, 6)
A or B or C full scale 3.
AIPM
- a. tWie
$.58(W-da) + 50%
X 3
1 (Note 1, 6, 7)
D or E (flow ref-(Note 2)
[
erenced)
- s k
b.
Downscale
> 3/125 full scal <
X 3
1 (Note 1, 6, 7)
D or E 5
E.
3.2/4.2 56 D
4 Table 3.2.3 - Contiraod Instninentation 11nt Initiates Rod Block
~
lbtes:
(1) "Ikre shall be tu) operable or operatirg trip systans for each furtion. If tie minhasa rud>er of operable or operatirg instninent channels camot be net for one of the tu) trip systam, this condition nay exist tp to seven days provided that Wrirg this tine tie operable systan is functionally tested innediately and daily threafter.
(2)
"W" is the reactor recirculation drivirg flow in percent, & - 0 for tu) recirculation loop operation, & - 5.4 for single recirculation loop operation.
(3) Only one of tk four SRM damels ng be bypassed.
(4) 'Ihere nust be at least one operable or operatirg IIM damel monitorirg each core quadrant.
(5) An RBMchamel will be considered imperable if tiere are less than half the total rudaer of nonnal irputs.
(6) 14mn discovery that mininten requironents for the rudaer of operable or operatirg trip systans or instnenent chamels are mt satisfied xtions shall be initiated to:
(a) Satisfy tie requirments by placirg appropriate dumels or systans in tk tripped condition or (b) Place tie plant mder the specified required conditions usirg runnal operatirg proce&res.
(7) 'lkre nust be a total of at least 4 operable or operatirg AHM dkurels (8) There are 3 tpscale trip levels. Ortly one is applied over a specified operatirg core tiennal power range. All RBM trips are autanatically bypassed below 30% tkrnal power.
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3.2/4.2 58 D
i 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.
When irradiated fuel is in the reactor vessel and regetor coolant temperature is less than 212 F, all low pressure core and containment cooling subsystems may be inoperable provided no work is being done which has the potential for draining the reactor vessel except as allowed by specification 3.5.C.4 below.
4.
When irradiated fuel is in the reactor vessel and the vessel head is removed, the suppression chamber may be drained completely and no more than one control rod drive housing or instrument thimble opened at any one time provided that the spent fuel pool gates are open and the fuel pool water level is maintained at a level of greater than or equal to 33 feet, i
a 8,
i c
3.5/4.5 113
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
- 11. Recirculation System II. Recirculation System i
- 1. The reactor may be started and operated, or
- 1. See Specification 4.6.G operation may continue with only one recirculation loop in oper4 tion provided
- 2. The followin$ baseline noise levels that:
will be obtained prior to operation with only one recirculation pump in operation at
- a. The following changes to setpoints and a core thermal power greater than that -
safety limit settings will be made within specified in Figure 3.5.1 or with a core 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiating operation with flow greater than 45% provided that only one recirculation loop in operation.
baseline values have not been established since the last core refueling. Baseline
will values will be taken with only one be changed per Specification 3.
recirculation pump running.
- 2. The Maximum Average Planar Linear lleat
- a. Establish a baseline core plate AP noise Generation Rate (M PL11GR) will be level.
changed as noted in Table 3.11.1.
Rod Block setpoints will be changed as noted in Specification 2.3.A and Table
- 3. With only one recirculation loop in 3.2.3.
operation at a core thermal power greater than that specified in Figure 3.5.1 or with 4
- b. Total core flow will f.e maintained a core flow greater than 45%, determine the
?l Sreater than 39% when core thermal power following noise levels at least once per 8 1s above the limit specified in Figure hour period and within 30 minutes after a 3.5.1.
core thermal power increase of greater than 5% of rated thermal power.
Prior to continued operation with onl
- c. one recirculation pump in operation, y
- a. Core plate A P noise levels.
- 1. the surveillance -requirements of
Specification 4.5.11.2 shall be met within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or, N
m
- 2. action shall be taken to:
I
- b. reduce total core flow to less than 45% and,
).
b.
reduce core thermal power to less i
o than the limit specified in Figure 3.5.1.
D 1.5/4.5 114 0
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS d.
If the core plate A P noise level is found to be greater than l.0 psi and 2 times its established baraline during the performance of Specification 4.5.H.3, immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by decreasing core flow and/or initiating an orderly reduction of core thermal power 4
by inserting control rods.
noise levels ar/or LPRM neutron flux If the APP.M and e.
found to be greater than e
three times their established baseline values during the performance of Specification 4.5.H.3, immediately initiate correct ve action to restore the i
noise levels to within the required a
limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow and/or initiating an orderly reduction of core thermal power by inserting control rods.
1 i
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2
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1.5/4.5 ll4a a
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80 70 f
60
/
E l
I 50 5
40 v
~
E fi!
30
{
CTP=(0.637)ICF+19.2 v
CTP= Core Thermal Power (% of 20 Rated)
ILF= Indicated C' ore Flow (% of Rated) 10
- 0. q '
L 0
30 40 50 60 70 80 Indicated Core Flow (% of Rated)
Figure 3.5.1 Single Loop Operation Surveillance Power / Flow Curve 3.5/4.5 ll4D Amendment No. 47 7
-~
-. ~
4 Bases Continued 3.5:
l C.
Emergency Cooling Availability The p'uypose of Specification G is to assure that sufficient core cooling equipment is available at all times.
It is during refueling outages that major maintenance is performed and during such time that all core and containment cooling subsystems may be out of service.
Specification 3.5.G.3 allows ' all i
core and containment cooling subsystems to be inoperable provided no work is being done which has the potential for draining the reactor vessel.
Thus events requiring core cooling are precluded.
4 i
Specification 3.5.G.4 recognizes that concurrent with control rod drive maintenance during the refueling outage, it may be necessary to drain the suppression chamber for maintenance or for the inspection required by Specification 4.7JLl.
In this situation, a sufficient inventory of water is maintained to assure adequate core cooling in the unlikely event of loss of control rod drive housing or instrument thimble seal integrity.
i H.
Recirculation System l
Specification 3.5.H.1 is based upon providing assurance that neutron flux limit cycle oscillations, which have a small probability of occurring in the high p/ low / low flow corner of the operating domain, are ower detected and suppressed. Under certain high power flow conditions that could occur during a recirculation pump trip and subsequent Single Loop Operation (SLO) where reverse flow occurs in inactive jet pumps, a hydraulic / reactor kinetic feedback mechanism can be enhanced such that sustained limit cycle oscillations of flow noise with peak to peak levels several times normal values are i
exhibited. Although large margins to safety limits are maintained when these limit cycle oscillations occur,by inserting rods and/or increasing coolantthey are to be monitored for, and suppressed when flux noise exceeds the three time b i
value flow. The line in Figure 3.5.1 is based on the 80%
l rod line below which the probability of limit cycle oscillations occurring is negligible, t
i APRM and/or LPRM oscillations in excess of those specified in Specification 3.5.H.l.e could be an I
indication that a condition of thermal hydraulic instability exists and that appropriate remedial l
action should be taken. By restricting core flow to greater than or equal to 39% of rated which corresponds to the core flow at the 80% rod line with 2 recirculation pumps running at minimum speed, the region of the power / flow map where these oscillations are most likely to occur is avoided (Ref.1).
~
Above 45% of rated core flow in Single Loop Operation there is the potential to set up high flow-i induced noise in the core.
Thus, surveillance of core plate P noise is rcquired in this region of the power / flow map to alert the operators to take appropriate remedial action if such a condition exists.
f Specification 3.6JL2 governs the restart of the pumbx transients and/or thermal stresses.
in an idle recirculation loop. Adherence to this specification limits the probability of excessive fi j
g_
I.
Deleted
?
jf Re ferences:
- 1. General Electric Service Information Letter No. 380, Rev. 1, February 10, 1984 i
i l
4, 3.5 BASES 119 1
%J i
i 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.11 REACTOR FUEL ASSEMBLIES 4.11 REACTOR FUEL ASSEMBLIES Applicability Applicability i
The Limiting Conditions for Operation The Surveillance Requirements apply to the associated with the fuel rods apply to those parameters which monitor the fuel rod parameters which monitor the fuel rod operating conditions.
}
operating conditions.
Objective Objective i
l The objective of the Limiting Conditions for The objective of the Surveillance i
@eration is to assure the performance of Requirements is to specify the type and-I the fuel rods.
frequency of surveillance to be applied to the fuel rods.
Specifications Specifications A. Average Planar Linear Heat Generation i
Kate (AFLHGK)
A. Average Planar Linear Heat Generation Kate (AFLHGK)
During power operation the APLHGR for l
all core locations shall not exceed the The APLHGR for each type of fuel as a i
appropriate APLHGR limit for those core function of average planar exposure shall locations. The APLHGR limit which is a be determined daily during reactor function of average planar ex,posure and operation at >25% rated thermal power.
I fuel twe is the appropriate value from j
Table 3.11.1 (based on a straight line interpolation between data points) for two recirculation loop operation, or 85%
4 of the appropriate value from Table 3.11.1 for one recirculation loop operation, I
multiplied by the smaller of the two MAPFAC factors determined from Figures 3.11.1 and 3.11.2.
If any time during operation it 3
Ef is determined that the limit for APLHGR is being exceeded action shall be t
m i
EL initiated within 15 a
i EF S$
i i
?'
3.11/4.11 211 a
%J
._.m
~
l j
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS C. Minimum Critical Power Ratio (MCPR)
C. Minimum Critical Power Ratio (MCPR) j If thermal power is greater than 45%, the MCPR shall be determined daily during' MCPR limit is the greater of:
reactor power operatica at >25% rated thermal power and following any change in
- 1) MCPR (100) from Table 3.11.2 multiplied power level or distribution which has the~
by from Figure.3.11.3 or, potential of brin ng the core to its operating MCPR Li t.
- 2) MCPR from Figure 3.11.4.
p If thermal power is less than or equal to 45%, the MCPR limit is obtained frc:
Figure 3.11.3.
The OLMCPR limit for one recirculation loop operation is 0.01 higher than the i
comparable two loop value.
If at any time during operation it is determined that the limiting value for MCPR is being exceeded, action shall be i
initiated within 15 minutes to restore operation to within the prescribed limits, j
Surveillance and corresponding action j
shall continue until reactor operation is within the prescribed limits.
If the j
steady state MCPR is not returned to j
within the prescribed limits within two (2) hours, the reactor shall be brou t to 4
the Cold Shutdown condition within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
YY 8
an b
I 3.11/4.11 213 a
%4 i
TABLE 3.11.1 MAXIMUM AVERAGE PIANAR LINEAR EEAT GENEPATION RATE vs. EXPOSURE Exposure MAPLHGR FOR EACH FUEL TYPE (kw/ft) 8DB262 MWD /STU 8DB250 8DRB282 P8DRB265L P8DRB282 P8DRB284LB P8DRB299L 8DB219L 8DRB265L BP8DRB265L BP8DRB282L BP8DRB284LB BP8DRB299L 200 11.1 11.2 11.6 11.2 11.4 11.0 1,000 11.3 11.2 11.6 11.2 11.4 11.0 5,000 11.9 11.6 11.8 11.8 11.8 11.6 10,000 12.0 11.7 11.9 11.9 11.9 11.9 15,000 11.9 11.7 11.9 11.8 11.9 11.9 20,000 11.8 11.5 11.8 11.7 11.7 11.8 25,000 11.3 11.3 11.3 11.3 11.4 11.5 30,000 10.2 10.7 10.7 11.1 10.8 10.9 35,000 9.6 10.2 10.2 10.4 10.2 10.3 m
R 40,000 8.9 9.6 9.6 9.8 9.5 9.7 45,000 8.9 9.0 Note: For two recirculation loop operation.
For single loop operation multiply these values by 0.85.
l r
w P
3.11/4.11 214 s
O
l Bases Continued This limit was determined based ugR operating limits for off-rated congitions is presentegiven core gowerin NEDC-3 492-P.(y) on bounding analyses for the limitin transient at the level.
Further information on MC At thermal power levels less than or equal to 25% of rated thermal power, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin. MCPR evaluation below this power level is therefore unnecessary. The daily requirement for calculating MCPR above 25% of rated thermal power is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
Those abnormal operational transients, analyzed in FSAR Section 14.5, which result in an automatic reactor scram are not considered a violation of the LCO.
Exceeding MCPR limits in such cases need not be reported.
Referencas
- 1. " Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (AR13) program for Monticello Nuclear Generating Plant", NEDC-30491-P, April, 1984.
- 2. " Analytical Methods of Plant Transient Evaluations for the GE BWR", NEDO-10802, February,1973.
- 3. " Response to NRC Request for Information on ODYN Computer Code", R H Bucholz to P S Check (USNRC),
September 28, 1977
- 4. " General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K", NEDE-20566, November 1975.
- 5. " Revision of Low Core Flow Effects on LDCA Analysis for Operating BWRs", R L Gridley (CE) to D C Eisenhut (USNRC), September 28, 1977
- 6. " Loss-of-Coolant Accident Analysis Report for the Monticello Nuclear Generating ) Plant",to Director of Nuclear Reactor Regulatio NEDO-24050-1 l
December, 1980, L 0 Mayer (NSP) 1 j
- 7. "Monticello Nuclear Generating Plant Single-Loop Operation", NEDO-24271, July,1980 i
Bases 4.11
{
EI The APLilGR, LilGR and MCPR shall be checked daily to determine if fuel burnup, or control rod movement have i
causedchankesinpowerdistribution. Since changes due to burnup are slow
- and only a few control rods are o.
removed dai y, a daily check of power distribution is -adequate.
For a limiting value to occur below 25% of i
rated thermat power an unreasonably large peaking factor would be required, hanges in the core power level or which is not the case for operating control ro,d seguences.
In addition, the MCPR is checked whenever c j
s j
distribution are made which have the potential of bringing the fuel rods to their thermal-hydraulic limits.
I
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o 45
'd 4.11 BASES 218 j
i 1
A
- .