ML19294A428
| ML19294A428 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/02/1978 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811130203 | |
| Download: ML19294A428 (16) | |
Text
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Wisconsin Electnc eowca coursur 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 November 2, 1978 CERTIFIED MAIL Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.
20555 Attention: Mr. A. Schwencer, Chief Operating Reactor Branch #1
Dear Mr. Denton:
DOCKET NOS. 50-266 AND 50-301 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 56_
OVERPRESSURE PROTECTION SYSTEM POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 In accordance with Part 50.59 of Title 10 of the Code of Federal Regula-tions, Wisconsin Electric Power Ccmpany (Licensee) hereby requests amendments to Facility Operating Licenses DPR-24 and DPR-27 to incorporate the following changes to the Technical Specifications for the Point Beach Nuclear Plant Units 1 and 2.
The Technical Specification changes discussed herein include measures to ensure the operability of the overpressure protection system and to impose limitations on operation of equipment which could exceed the basis of design for the overpressure protection system.
Licensee was requested to submit these Technical Specification changes by Mr. Schwencer's letter dated September 11, 1978.
In addition to these changes, we are takir.2 this opportunity to submit for your approval a number of administrative changes to the Technical Specifications.
These changes are discussed in greater detail below.
A new Specification 15.3.15, " Overpressure Mitigating System Operation" has been provided herewith.
The purpose of this specification is to specify the functional requirements and limiting conditions for operatiois concerning the use of the pressurizer power operated relief valves as part of the overpressure mitigating system at the Point Beach Nuclear Plant.
(bis specification also includes limita-tions on the operability of the hign pressure safety injection system when the reactor coolant system is at low temperatures and restrictions on starting a reactor coolant pump when the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test. An addition has been made to Table 15.4.1-1 to specify the check, calibration and testing requirements for the over-pressure mitigating system. A special reporting requirement has been added to Specification 15.6.9.3 which requires a report to the NRC in the event the over-pressure protection system is operated to relieve an overpressure transient.
.g 18111
Mr. Harold R. Denton - Page Two November 2, 1978 As mentioned previously, Licensee is also including with this request a number of minor specification changes which are administrative in nature.
These changes are summarized as follows:
a.
A correction to the last column heading in Tables 15.3.5-2, 15.3.5-3 and 15.3.5-4.
This heading should read " Operator Action If Conditions Of Column 3 or 4 Cannot Be Met".
The necessity for this chang'e has been pointed out by NRC inspectors several times.
b.
Deletion of the column headed "FSAR Section Reference" for Table 15.4.1-2.
It has been pointed out by NRC inspectors that many of the references given are no longer correct.
Since such references are not required in the Specifications, we suggest that they be deleted.
c.
The comment concerning well water samples in Table 15.4.10-1 has been deleted.
This comment was not applicable to well water samples and was originally inserted due to a clerical error.
d.
A change has been propored to Specification 15.4.15.F.2 to delete the year reference in the diesel fuel sampling and acceptance procedures.
Diesel fuel oil testing is performed in accordance with the latest approved ASTM test procedures.
e.
Specification 15.6.11 has been changed to correct the title of the referenced manual.
You have informed us that the NRC has determined this license amendment to be a Class III Amendment.
Since this license amendment request involves twa units and the charges requested for both units are identical, we have determined that the amendment for the second unit is a Class I Amendment. Accordingly, we have enclosed herewith check number 407828 in the amount of $4,400 for the amendment fees.
We have provided herewith three signed originals of this change request.
We shall forward under separate cover forty copies of the request.
This request includes the attached proposed ravised Technical Specification pages which detail the changes discussed herein.
Should you have any questions concerning these changes, please contact us.
Very truly yours,
/
n o Sol Burstein Executive Vice President Enclosures Subscribed cad sworn to before me this 2nd day of November 1978.
thWYi N
[x-Notary Publ $, State of Wisconsin My Commission expires M ( / 97o.
A A
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TABLE 15.3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP 1
2.
3 4
5 NO. OF MIN.
MINIMUM PERMISSABLE OPERATOR ACTION NO. OF CHANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF NO.
FUNCTIONAL UNIT CRANNELS TO CRANNELS REDUNDANCY CONDITIONS COLUMN 3 OR 4 TRIP CANNOT BE MET 1.
Manual 2
1 1
Maintain hot shutdown 2.
Nuclear Flux Power Range **
2 of 4 power range Maintain hot low setting a
2 3
2 channels greatur than shutdown high setting 4
2 3
2 10% F.P.
(low setting only) 3.
Nuclear Flux Intermediate Range 2
1 1
2 of 4 power range channels Maintain hot greater than 10% F.P.
shutdown. Note 1 4.
Nuclear Flux Source Range 2
1 1
1 of 2 intermediate rangg10 "I" ^ "
channels greater than 10 shutdown. Note 1 amps 5.
Overtemperature AT 4
2 3
2 Maintain hot shutdown 6.
Overpower AT 4
2 3
2 Maintain hot shutdown 7.
Low Pressurizer Pressure 4
2 3
2 Maintain hot shutdown 8.
Hi Pressurizer Pressure 3
2 2
1 Maintain hot shutdown 9.
Pressurizer-Hi Water Level 3
2 2
1 Maintain hot shutdown
- 10. Low Flow in one loop (>50%
3/ loop 2/ loop 2
1 Maintain hot F.P.)
(any loop) shutdown Low Flow Both Loops (10-50%
3/ loop 2/ loop F.P. )
(any loop)
TABLE 15.3.5-2 (Cont 'd) 1 2
3 4
5 NO. OF MIN.
MINDtUM PERMISSIBLE OPERATOR ACTION NO. OF C11ANNELS OPERABLE DEGREE OF BYPASS IF CONDITIONS OF NO.
FUNCTIONAL UNIT QIANNELS TO CHANNELS REDUNDANCY CONDITIONS COLUMN 3 OR 4 TRIP CANNOT BE NET 11.
2 2
1 Maintain <50% of rated power 12.
Steam Flow - Feed Water Flow 2/ loop 1/ loop 1/ loop 1/ loop Maintain h'ot mismatch shutdown
^
13.
Lo Lo Steam Generator 3/ loop 2/ loop 2/ loop 1/ loop Maintain hot Water Level shutdown Maintain hot 14.
Undervoltage 4 KV Bus 2/b us 1/ bus 1/bes (both buses) shutdown Maintain hot 15.
Underfrequency 4 KV Bus 2/ bus 1/ bus 1/ bus (both buses) shutdown 16.
Control rod misalignment as 1
1 Log individual rod monitored by on-line computer positions once/ hour, and after a load change
>10% or after >30 inches of control rod motion NOTE 1: When block condition exists, maintain normal operation.
Full Power F.P.
=
- Not Applicable
- One additional channel may be taken out of service for zero power physics testing.
O V
U
O O
O TABLE 15.3.5-3 EMERGENCY COOLING
.c 1
2 3
4 5
OPERATOR ACTION NO. OF MIN.
MIN.
PERMISSIBLE IF CONDITIONS OF NO. OF CHANNELS OPERABLE DEGREE OF DYPASS COLU!Of 3 OR 4 NO.
FUNCTIONAL UNIT CHANNELS TO TRIP CHANNELS REDUNDANCY CONDITIONS CANNOT BE MET 1.
SAFETY INJECTION a.
Manual 2
1 1
1 Hot Shutdown ***
b.
High Containment Pressure 3
2 2
1 Hot Shutdown ***
c.
Steam Generator Iow Steam Primary Pressure is 14 Pressure / Loop 3
2 2
1 Less than 1800 psig Hot Shutdown ***
1 d.
Pressurizer Low Pressure Primary Pressure is and Iow Level 3*
1*
1*
2 Less than 1800 psig Hot Shutdown ***
2.
Manual 2
2 2
Hot Shutdown ***
b.
Hi-Hi Containment Pressure 2 sets 2 of 3 2 per 1/ set Hot Shutdown ***
(Containment Spray) of 3 in each set set
- - Each Channel has one pressurizer pressure and one pressurizer level signal.
- - Must actuate 2 switches simultaneously.
- - If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken on the affected unit to place the unit in cold shutdown conditions.
b.
TABLE 15. 3. 5-4 INSTRUMENT OPERATING CONDITIONS FOR ISOLATION FUNCTIONS 1
2 3
4 5
MINIMUM PERMISSABLE NO. OF MIN.
DEGREE BYPASS OPERATOR ACTION NO. OF CHANNELS OPERABLE OF CONDITIONS IF CONDITIONS OF NO.
FUNCTIONAL UNIT CHANNELS TO CHANNELS REDU-COLUMN 3 OR 4 TRIP DANCY CANNOT BE MET 1 CONTAINMENT ISOLATION a.
Safety Injection See Item No. 1 b,c, and d of Hot Shutdown ***
Table 15.3.5-3 b.
Manual 2
1 1
Hot Shutdown 2 STEAM LINE ISOLATION a.
Hi Hi Steam Flow with 2/ loop 1
1 Safety Injection Hot Shutdown ***
b.
Ili Steam Flow 2/ loop I
1 Hot Shutdown ***
and 2 of 4 Low T with Safety InjeSEEon c.
Hi Containmen*. Pressure 3 2
2 1
Hot Shutdown ***
d.
Manual 1/ loop 1/ loop 1/ loop Hot Shutdown
- If minimum conditions are not met within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, steps shall be taken on the affected unit to place the unit in cold shutdown conditions.
15.3.15 Overpressure Mitigating System Operations Applicability Applies to operability of the overpressure mitigating system when the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test.
Ok,iective To specify functional requirements and limiting conditions for opt.ation on the use of the pressurizer power operated relief valves when used as part of the overpressure mitigating system and to specify further limiting conditions for operation when the reactor ' coolant system is operated without a pressure absorbing volume in tae pressurizer.
Specification A.
System Operability 1.
Except as specified in 15.3.15.A.2 below, the overpressurization miti-gating system shall be operable wl.enever the reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressurization temperature for the inservice pressure test, as specified in Figures 15.3.1-1 (Unit 1) and 15.3.1-3 (Unit 2). Operability require-ments are:
a.
Both pressurizer power operated relief valves operable at a setpoint of 1425 psig, b.
The upstream isolation valves to both power operated relief valves are open.
2.
The requirements of 15.3.15.A.1 may be modified to allow one of the two power operated relief valves to be inoperable for a period of not more than seven days.
15.3.15-1
3.
If the inoperable power operated relief valves cannot be made operable within seven days, the reactor coolant system must be depressurized and vented to the pressurizer relief tank within eight hours.
4.
If both power operated relief valves are inoperable, the reactor coolant system must be depressurized and vented to the pressurizer relief tank within eight hours.
B.
Additional Limitations 1.
When the reactor coolant system is not open to the atmosphere and the temperature of one or both reactor coolant system cold legs is <275 F, a
no more than one high pressure safety injection pump shall be operable.
The second high pressure safety injection pump shall be demonstrated inoperable whenever the temperature of one or both reactor coolant system cold legs is <275 F by verifying that the motor circuit breakers have been removed from their electrical power supply circuits or by verifying that the discharge valves from the high pressure safety injection pumps to the reactor coolant system are shut and that power is removed from their operators.
2.
A reactor coolant pump shall not be started when the reactor coolant system temperature is less than the minimum temperature for the inservice pressure test unless:
a.
There is a pressure absorbing volume in the pressurizer or in the steam generator tubes or b.
The secondary water temperature of each steam generator is less than 50*F above the temperature of the reactor coolant system.
Basis The Overpressurization Mitigating System consists of a diverse means of relieving pressure durir.g periods of water solid operation and when the system temperature is below the value pemitted to perform the primary system leak test.
This method 15.3.15-2
of water relief utilizes the pressurizer power operated relief valves (PORV's).
The PORV's are made operational for low pressure relief by utilizing a dual set-point where the low pressure circuit is energized and de-energized by the operator with a keylock switch depending on plant conditions. The logic required for the low pressure setpoint is in addition to the existing PORV actuation logic and will not interfere with existing automatic or manual actuation of the PORV's.
During plant cooldown prior to reducing reactor coolant system temperature below the minimum temperature allowable for the inservice pressure test, the operator under administrative procedures shall place the keylock switch in the " Low Pressure" position. This action enables the Overpressure Mitigating System.
The redundant PORV channels shall remain enabled and operable while the reactor coolant system is not open to the atmosphere and the temperature is less than the minimum pressuri-zation temperature for the inservice pressure test, except that one PORV may be out of service for a period of up to seven days.
t T? a mass input transient used to determine the PORV setpoint assumed a worse case transient of a single high pressure safety injection pump discharging to the reactor coolant system while the system is solid. Therefore, when the reactor coolant system is less than 275 F, only one high pressure safety injection pump shall be operable at any time except when the reactor coolant system is open to the atmosphere.
The heat input transient used to detennine the PORV setpoint assumes a temperature difference between the reactor coolant system and the steam generator of 50 F.
Therefore, before starting a reactor coolant pump when the reactor coolant system is solid, the operator shall insure that the secondary temperature of each steam generator is less than 50 F above the temperature of the reactor coolant system unless a pressure absorbing volume has been verified to exist in the pressurizer or steam generator tubes.
15.3.15-3
TABLE 15.4.1-1 (CONTI!."JED) t Channel Description Check Calibrate Test Remarks 24.
Containment Pressure S
R M**
Narrow range containment pressure
(-3.0, +3 psig excluded) 25.
Steam Generator Pressure S***
R M***
26.
Turbine First Stage Pressure S**
R M**
27.
Emergency Plan Radiation M
R M
Instruments 28.
Environmental Monitors M
N.A.
N.A.
29.
Overpressure Mitigating System S
R S
Each Shift M
Monthly D
Daily P
Prior to each startup if not done previous week Each Refueling Shutdown (But not to exceed 20 months, W
Weekly R
except for first core cycle)
B/W -
Biweekly NA -
Not applicable Not required during periods of refueling shutdown, but must be performed prior to starting up if it has not been performed during the previous surveillance period.
Not required during periods of refueling shutdown if steam generator vessel temperature is greater than 70'F.
Each PORV shall be demonstrated operable by:
Performance of a channel functional test on the PORV actuation channel, but excluding valve operation, a.
within 31 days prior to entering a condition in which the PORV is required operable and at least once per 31 days thereafter when the PORV is required operable.
TABLE 15.4.1-2 MINIMUM FPIQUENCIES FOR EQUIPMENT AND SAMPLING TESTS Test Frequency 1.
Reactor Coolant Samples Gross Beta-gamma 5/ week (7) activity (excluding tritium)
Tritium activity Monthly Radiochemical E Semiannually (2)
Determination (1)
Chloride Concentration 5/ week (8)
Diss. Oxygen Conc.
5/ week (6)
Fluoride Conc.
Weekly j
~
2.
Reactor Coolant Boron Boron Concentration Twice/ week 3.
Refueling Water Storage Boron Concentration Weekly (6)
Tank Water Sample 4.
Boric Acid Tanks Boron Concentration Twice/ week 5.
Spray Additive Tank NaOH Concentration Monthly 6.
Accumulator Baron Concentration Monthly 7.
Spent Fuel Pit Boron Concentration Monthly 8.
Secondary Coolant Gross Beta-gamma acti-Weekly (6) vity or gamma isotopic analysis Iodine concentration Weekly when gross Beta-gamma activity equals or exceeds 1.2 pCi/cc (6) 9.
Control Rods Rod drop times of all Each refueling or full length rods (3) after maintenance that could affect proper functioning (4) 10.
Control Rod Partial movement of Every 2 weeks (6) all rods 11.
Pressuriner Safety Valves Set point Each refueling shutdown 12.
Main Steam safety Valves Set point Each refueling shutdown 13.
Containment Isolation Trip Functioning Each refueling shutdown
TABLE 15.4.1-2 (CONTINUED)
T_est Frequency e
14.
Refueling System Interlocks Functioning Each refueling shutdown 15.
Service Water System Functioning Each refueling shutdown 16.
Primary System Leakage Evaluate Monthly (6) 1/.
Diesel Fuel Supply Fuel inventory Daily 18.
Turbine Stop and Functioning Monthly (6) l9)
Governor Valves 19.
Low Pressure Turbine Visual and magnetic Every five years Rotor Inspection (5) particle or liquid j
penetrant 20.
Boric Acid System Storage Tank Daily Temperature 21.
Boric Acid System Visual observation Daily of piping temperatures (all >145'F) 22.
Boric Acid Piping Heat Electrical circuit Monthly Tracing operability
\\
(1)
A radiochemical analysis for this purpose shall consist of a quantative measure-ment of each radionuclide with half life of >30 minutes such that at least 95%
of total activity of primary coolant is accounted for.
(2)
E determination will be started when the gross activity analysis of a filtered sample indicates >10 pc/cc and will.be redetermined if the primary coolant gross radioactivity of a filtered sample increases by more than 10 pc/cc.
(3)
Drop tests shall be conducted at rated reactor coolant flow.
Rods shall be dropped under both cold and hot conditions, but cold drop tests need not be time.
(4)
Drop tests will be conducted in the hot condition for rods on which maintenance was performed.
(5)
As accessible without disassembly of rotor.
(6)
Not required during periods of refueling shutdown.
(7)
At least once per week during periods of refueling shutdown.
(8)
At least three times per week (with maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples) during periods of refueling shutdown.
(9)
Effective in March 1978, the requirement for the monthly function test for Point Beach Unit No. 2 is waived until the start of the fourth refueling outage.
TABLE 15.4.10-1 (CONTINUED)
Sample Type Locations (a,c)
Frecuency Analysis Comments s
Well Water 1-Onsite Well (10)
Quarterly Gross Beta-T.S. (b)
Gamma Scan T.S.
Tritium Strontium-89 Strontium-90 Milk 1-Iccal dairy pool (11)
Monthly Gamma Scan Radioiodine analysis 1-Dairy Farm, NNW (19)
Radiciodine done by the resin 1-Dairy Farm, SSE (21)
Strontium-89 extraction technique.
Strontiur - 9p Algae 1-North of Discharge (5) 3x/yr Gross Beta 1-Discharge of Flume (12) as available Gamma Scan Fish 1-Travelling screens (13) 3x/yr Gross Beta Analysis of edible as available Gamma Scan portions only.
(a)
Reference location is chosen well in excess of 10 miles from the plant in a low X/Q sector to provide an estimate of background levels.
(b)
T.S. - Total Solids (c)
Numbers given under location correspond to sampling locations shown in Figure 15.4.10-1.
Page 2 of 2
C.
Fire Hose Stations Test Frequency l.
Visual Inspection Monthly 2.
Hose Hydro-Test Yearly 3.
Partially open each hose station valve 3 years to verify operability and no blockage D.
Fire Detection Test Frequency 1.
Channel Functional Test 2 mo.
E.
Fire Barrier Penetration Fire Seals Test Frequency J
~
1.
Visual Inspection 18 mo. and following repairs or maintenance F.
Fire Pump Diesel Engine Test Frequency 1.
a.
Verify 200 gallons of fuel in Monthly fuel storage tank b.
Verify diesel starts from ambient Monthly conditions and operates for at least 20 minutes.
2.
Sample diesel fuel per ASTM-D270 Quarterly and verify acceptable per Table 1 of ASTM-D975 with respect to viscosity, water content and sediment.
3.
a.
Inspect diesel per procedures 18 months prepared in conjunction with its manufacturer's recommendations b.
Verify diesel starts from ambient 18 months conditions and operates for >20 minutes while loadqd with the fire pump 15.4.15-2 9-
(1) The number and types of samples taken and the reasurements made on the samples; e.g., gross beta gamma scan, etc.
(2) Any changes made in sample types or locations during the reporting period, and criteria for these changes.
b.
A summary of survey results during the reporting period.
4.
Leak Testing of Sources Results of required leak tests performed on seal sources if the tests reveal the pressure of 0.005 microcuries or more of removable contamination.
D.
Overpressure Mitigating System Operation In the event the overpressure mitigating system is operated to relieve a pressure transient which, by licensee's evaluation, could have resulted in an overpressurization incident had the system not been operable; a special report shall be prepared and subm.tted to the Commission within 30 days. The report shall describe the circumstances
\\
initiating the transient, the effect of the system on the transient and any corrective action necessary to prevent recurrence.
15.6.9-10
15.6.11 RADIATION PROTECTION PRO RAM Specification Radiological control procedures shall be written and made available to all station personnel, and shall state permissible radiation exposure icvels. The radiation protection program shall meet the requirements of 10 CFR 20, with the exception of the following:
Paragraph 20.203 - Caution signs, labels and signals In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c) (2), each radiation area in which the intensity of radiation is greater than 100 mrem /hr shall be barripaded and conspicuously posted as a High Radiation Area, and entrance thereto shall be controlled in accordance with the Point Beach Nuclear Plant Health Physics Administrative Control Policies and Procedure Manual, Section 2.7, Radiation Work Permit. A person er persons permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
In addition, each High Radiation Area outside the containment building in which the intensity of radiation is greater than 1000 mrem /hr shall be provided with locked barricades to prevent unauthorized entry into these areas, and the keys to these locked barricades shall be maintained under the administrative control of the Duty Shift Supervisor.
15.6.11-1
,