ML19332G045
| ML19332G045 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/07/1978 |
| From: | Lear G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19332G043 | List: |
| References | |
| NUDOCS 8912200073 | |
| Download: ML19332G045 (12) | |
Text
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UNITEo STA788 y
NUCLEAR REGULATORY 00MMIS$10N s
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wasvaworow, o.c. seems OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 i
License No. DPR-40 1.
The Nuclear Regulatory Commission (the Commission) has found that:
j A.
The application for amendment by Omaha Public Power District (the licensee) dated Augusi! 5,1977, complies with the standards and requirements of tne Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR y
Chapter I; 1
i The facility will operate in conformity with the application, B.
the provisions of the Act, and the rules and regulations of
.the Commission; C.
There is reasonable assurance (i) that the activities authorized l
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 Commission's regulations;
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D.
The issuance of this amendment will not be inimical to the l
common defense and security or to the health and safety of I'
the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part l
51 of the Commission's regulations and all applicable requirements l
have been satisfied.
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p 8912200073 780307 PDR ADOCK 05000205 19 R&L
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L 2.
Accordingly, the license is amended by changes to the Technical l
Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License L-No. DPR-40 is hereby amended to read as follows:
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(B) Technical Specifications l'
The Technical Specifications contained in Appendices
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A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee-
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shall operate the facility in accordance with the Technical Specifications.
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3.
This license amendment is effective as of the date of its issuance.
'I FOR THE NUCLEAR REGULATORY COMMISSION a
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George LedT, Chief Operating Reactors Branch #3 Division of Operating Reactors L
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Attachment:
l Changes to the Technical l
Specifications Date of Issuance: March 7, 1978 I
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ATTACHMENT TO LICENSE AMENDMENT NO. 39 TO THE TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 l
Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised page is identified by Amendment number and contains vertical lines indicating the area of change.
Remove Replace 2-2 2-2 2-2a (new) l 2-15 2-15 2-16 2-16 2-22 2-22 2-23 2-23 2-23a (new) 3-16 3-16 3-16a (new) i
,+mm
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2.0 LIICTING CONDITIONS FOR OPERATION 2.1
[eactor Coolant System (Continued) i 2.1.1 Operable components (Continued)
(6)
Marimum steam generator steam side leak test pressure shall 1
not exceed 1000 psia. A minimum temperature of 820F is required.
i (9)
A non-operating reactor coolant pump shall not be started unless at least one of the following conditions is met:
(a)
A pressurizer steam space of 60% by volume or greater exists, or (b)
The steam generator secondary side temperature is less than 50 F above that of the reactor coolant system cold leg.
Basis When reactor coolant boron concentration is being changed, the pro-cess must be uniform throughout the reactor coolant system volume to prevent stratification of reactor coolant at lover boron concen-j tration which could result in a reactivity insertion. Sufficient mixing of the reactor coolant is assured if one low pressure safety injection pump or one reactor coolant pump is in operation. The i
low pressure safety injection pump will circulate the reactor coolant system volume in less than 35 minutes when operated at rated capacity.
The pressurizer volume is relatively inactive; therefore, it will tend to have a boron concentration higher than the rest of the re-actor coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pres 9 *izer and the reactor coolant system during the addition of boren.L Both steam generators are required to be filled above the lov steam generator water level trip set point wher.sver the temperature of the reactor coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.
The design cyclie transients for the reactor system are given in FSAR Section L.2.2.
In addition, the steam generators are designed for additional conditions listed in FSAR Section L.3.h.
Flooded and pressurized conditions on the steam side assure minimum tube sheet temperature differential during leak testing. The minimum temperature for pressurizing the steam generator steam side is 700F.
I Fonmation of a 60% steam space ensures that the resulting pressure increase vould not result in an overpressurization, should a re-actor coolant pump be started when the steam generator secondary p
side temperature is greater than that of the BCS cold leg.
l Amendment No. 39 2-2
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L2.0
'AMITING CONDITIONS FOR OPERATION' 2.1
- Waetor Coolant System (Continued)
- 2.1.1-Operable Components (Continued) 4 For the case in which no pressuriser steam space exists, limitation of the~ steam generator secondary side /RCS cold les AT-to 500F en-sures that a single low set point PORY would prevent an overpressuri-sation due to actuation of a reactor coolant puipp.
l References (1)
FSAR Section 4.3.7 i
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' Amendment No. 39
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- 9 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Resetor Coolant System (Continued) 2.1.6 Pressurizer and Steam System Safety Valves
_ Applicability Applies to the status of the pressurizer and steam system safety valves.
Objective To specify minimum requirements p<.rtaining to the pressurizer and steam system safety valves.
Specifications To provide adequate overpressure protection for the reactor coolant system and steam system, the following safety valve requirements shall be mett (1)
The reactor shall not be made critical unless the two pres-1 suriter safety valves are operable with their lift settings 2545 psia 1,1%.11) ensure valve opening between 2500 psia and sdjusted to (2)
Whenever there is fuel in the reactor, and the reactor vessel head is installed, a minimum of one operable safety. valve sha13 be installed on the pressurizer. However, when in at least the cold shutdown condition, safety valve nozzles may be open j
to containment atmosphere during performance of safety valve tests or maintenance to satisfy this specification.
(3)
Whenever the reactor is in power operation, eight of the ten steam safety valves shall be operable with their lift settings between 1000 psia and 1050 psia with a o erance of 1,1% of the nominal nameplate set point values.1 (h)
Both pressurizer power-operated relief valves (PORV's) shall be operable, at the low setpoint, whenever the cold leg temperature is less than 300'F. One PORV may be inoperable for up to 7 days, provide the remaining PORV is operable.
If the above conditions of this paragraph cannot be met, the primary system must be depressurized and vented.
Basis The highest reactor coolant system pressure reached in any of the accidents analyzed was 2k80 ps'ia and resulted from a complete loss of turbine generator load without simultaneous reactor trip while operating at 1500 MWt,(2) The reactor is assumed to trip on a "High Pressurizer Prescare" trip signal.
To determine the maximum steam flow, the only other pressure relieving Con-system assumed operational is the steam system safety valves.
servative values for all systems parameters, delay times and core moderator, coefficients are assumed. Overpressure protection is Amendment No. 39 2-15
.s-2.0 ' LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (Continued) 2.1. 6 Pressurizer and Steam System Safety Valves (Continued) provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and j
elevation heads.
If no residual heat were removed by any of the means available, the smount of steam which could be generated at safety valve lift pres-sure vould be less than half the capacity of one safety valve. This specification, therefore, provides adequate defense against over-pressurization when the reactor is suberitical.
Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere vill assure. that sufficient relief capacity is available.
Use of plastic l
or other similar material to prevent the entry of foreign material into the open nozzle vill not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.
The total relief capacity of the ten steam system safety valves is 6.5h x 106 lb/hr At the rated power of 1h20 MWt, a relief capacity of only k.7 x 10b lb/hr is required to prevent overpressurization of the steam syste= on loss-of-load conjitions and eight valves pro-vide relieving capability of k.976 x 100 lb/hr.(3)
Alignment of the power-operated relief valve lov setpoint below 3000F provides sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3. to prevent the design basis pressure transients from causing an overpressuriza-tion incident. Limitation of this requirement to scheduled cool-down ensures.that, should emergency conditions dictate rapid cool-down of the reactor coolant system, inoperability of the lov tem-perature overpressure protection system would not prove to be an inhibiting factor.
Removal of the reactor vessel head provides sufficient expansion volume to limit any of the design basis pressure transients.
- Thus, no additional relief capacity is required.
References (1)
Article 9 of the 1968 ASME Boiler and Pressure Vessel Code,Section III (2)
FSAR, Section 1h.9 (3)
FSAR, Sections k.3.h, k.3 9 5 Amendment No. 39 2-16
2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Baeratency Core Cooling System (Continued)
(3)
Wenever the reactor coolant system cold leg temperature is below 2100F and the reactor vessel head is insta?, led, at least
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two (2) HFSI pump control switches shall be pieced in pull-stop.
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Wenever the reactor coolant system cold leg tempcrature is below 1100F and the reactor vessel head is installed, all l
three (3) HPSI pump control switches shall be placed in pull-1 stop.
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1 In the event that no charging pumps are operable, a single RPSI pump may be taken from pull-stop and utilized for boric
.cid injection to the core.
Basis The normal procedure for starting the reactor is to first heat.the reactor coolant to near operating temperature by running the reactor coolant pumps. The reactor is then made critical by withdraving CEA's and diluting boron in the reactor coolant. With this mode of start-up, the energy stored in the reactor coolant during the approach to criticality is substantially equal to that during Power operation and therefore all enginened safety features and auxiliary cooling systems are required to be fully operable. Daring lov power Physics tests at lov temperatures, there is a negligible amount of stored energy in the reactor coolant; therefore, an accident com-Parable in severity to the design basis accident is not possible and the engineered safeguards systems are not required.
The SIRW tank contair, a min (mum of 283,000 gallons of usable water containing 1900 ppm torce.(ll This is sufficient boron concentra-tion to pro /ide a shutdown margin of 5%, including allowances for uncertainties, with al gentrol rods withdran and a new core at a temperature of 60cF. 2; Tne limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyses. The minimum 116.2 inch level corresponds to a volume of 825 ft3 and the maximum 128.1 inch level corresponds to a volume of 895.5 ft3 Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked.
Since the system is used for shut-down cooling, the valving vill be changed and must be properly aligned prior to start-up of the reactor.
The operable status of the various systems and, components is to be demonstrated by pcriodic tests. A large fraction of these tests vill be perfo~med while the reactor is operating in the power range.
2-22 Amendment !!o.
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I 2.0 LIMITING CONDITIONS FOR OPERATION 2.3 Dnergency Core Cooling System (Continued)
If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full oper-ability within a relatively short time. For a single component 1
to be inoperable does not negate the ability of the system to per-i form its function, but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failu es.
To provide maximum assurance that the redundant component (s) vill operate if required to do so, the redundant com-ponent(s) is to be tested prior to initiating repair of the inoper-able component.
If it develops that the inoperable component is j
not repaired within the specified allowable time period, or a second I
component in the same or related system is found to be inoperable, the reactor vill initially be put in the hot shutdown condition to provide for reduction of cooling requirements after a postulated loss-of-coolant accident. This vill also permit improved access for repairs in.some cases. After a limited time in hot shutdown, if the malfunction (s) is not corrected, the reactor vill be placed in the cold shutdown condition utilizing normal shutdown and cool-I down procedures.
In the cold shutdown condition, release of fission i
products or damage of the fuel elements is not considered possible.
The plant operating procedures vill require immediate action to effect repairs of an inoperable component and therefore in most cases repairs vill be completed in less than the specified allow-able repair times.
The limiting times to repair are intended to assure that operability of the component vill be restored promptly and yet allow sufficient time to effect repairs using safe and pro-per procedures.
The requirement for core cooling in case of postulated loss-of-coolant accident while in the Mt shutdown condition is significantly re-1 duced be1%' the requirements for a postulated loss-of-coolant acci-dont during power operation. Putting the7eactor in the hot shut-down condition reduces the consequences of a loss-of-coolant acci-dent and also allows more free access to some of the engineered safe-guards components in order to effect repairs.
Failure to complete repuirs within kB hours of going to the hot shut-down condition is considered indicative of a requirement for major maintenance and, therefore, in such a case, the reactor is to be put into the cold shutdown condition.
With respect to the core cooling function, ther9 s unctional re-dundancy over most of the range of break sizes.L3 (D The LOCA analysis confirns adequate core cooling for the break spectrum up to and including the 32 inch double-ended break assuming the safety injection capability which most adversely affects accident consequences and are defined as follows. The entire contents of all four safety injection tanks are assumed to be available for emergency core cooling.
but the content s of one of the tanks is assumed to be lost through the reactor coolant system.
In addition, of the three high-pressure Arendment No.
39
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2.0 pIMITING CONDITIONS FOR OPERATION 2.3 Deeraeney Core Cooling System (Continued) safety injection pumps and the two low-pressure safety injection pumps, for large break analysis it is assumed that two high pressure and one i
low pressure operate while 9nly one of each type is assumed to operate i
in the small break analysist5i; and also that 25% of their combined discharge rate is lost frem the reactor coolant system out of the break. The transient hot spot fuel clad temperatures for the break sites considered are shown on FSAR Figures 1-19 (Amendment No. 3k).
Placing at least two RPS7 pump control switches in pull-stop below j
210 F results in so more than one NPSI pump remaining operable.
A i
0 single low setpoint PORY is sufficient to prevent an overpressuri-sation, caused by operation of one RPSI pump and three charging pumps, above a cold leg temperature of 1100F. Placing of all three EPSI pump control switches in pull-stop below 1100F results in no RPSI pumps remaining operable. A single lov setpoint PORY is suffi-cient to prevent an overpressurization, caused by operation of three charging pumps, at any cold leg temperature.
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i Technical Specification 2.2(1) specifies that, when fuel is in the reactor, at least one flow path shall be provided for borie acid injection to the core.
Should boric acid injection become neces-sary, and no che,rging pumps are operable, operation of a single HPSI pump would provide the required flow path.
Referenees (1)
FSAR, Section ik.15 1 (2)
FSAR, Section 6.2.3.1 (3)
PSAR, Section 1k.15 3 (k)
FSAR, Appendix K (5)
Omaha Public Power tistrict's Submittal, December 1, 1976.
2-23a Amendment No. 39
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g TABLE 3-3 (Continued) 3 MINIMUM FREQUENCIES FOR CHECKS. CALIBRATIONS AND ',*ESTING OF MISCELIANEOU3 INGTRUMENTATION AND CONTROLS
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Surveillance Channel Description Function Frequency Surveillance Method Ih.
Nuclear Detector ' Jell a.
Test S
a.
Compare eight (8) indepenient Cooling Annulus Exit readings.
Air Te=perature Detectors b.
Calibrate R
b.
Calibrate with known temperature.
15 Reacter Coolant System a.
Check M
a.
Calculation of reactor coolant Flow flow rate.
w d.
16.
Pressurizer Pressure a.
Check S
a.
Cosaparison of independent pressure readings.
17.
Reactor Coolant Inlet a.
Check S
a.
Comparison of independent tempere-Temperature ture readings.
18.
Low-Temperature Set-a.
Test PM a.
Verify operability of actuation cir-point Power-Operated cuitry for low-temperature setpoint Relief Valves power-operated relief valves by utili-zation of installed test switches.
b.
Calibrate R
b.
Calibrate temperature and pressure channels.
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-k TABLE 3-3 (continued)
MINIMJM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING y
OF MISCELLANEOU3 INSTRUMENTATION AND CONTROLS
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U Surveillance Channel Description Function Frequency Surveillence Method S - Each Shift D - Daily M - Monthly 4
A - Annually gs R - 18 Months P - Prior to each startup if not perfonned within previous week.
PM - Prior to scheduled cold leg cooldown below 3000F; monthly whenever temperature remains below l
3000F and reactor vessel head is installed,
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