ML17250A779

Tech Specs Table of Contents,Page ii;1.0,definitions & 3.0 for Limiting Conditions of Operation

ML17250A779
Person / Time
Site:	Ginna
Issue date:	11/12/1980
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
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ML17250A777	List: ML17250A776 ML17250A778 ML17250A779
References
NUDOCS 8011200277
Download: ML17250A779 (27)

Text

Attachment A Remove Technical Specification pages 3 1 4 3 8 2 4 3 1 4 3 2 4 ll ii, 1-1, 1~ 4 11 2 3.1-1 through Insert revised Technical Specification pages ii, l-l 3.1-1 through 3.1-4, 3.1-4a, 3.1-4b, 3.1-4c, 3.8-2, 3.8-2a, 3.8-4, 3.11-2 through 3.11-4, 4.3-1 through 4.3-3, 4.11-1 through 4.11-3.

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TABLE OF CONTENTS (cont.)

Pacae 4.0 SURVEILLANCE REQUIREMENTS 4.1 Operational Safety Review 4.1-1 4.2 Inservice Inspection 4.2-1 4.3 Reactor Coolant System 4.3-1 Containment Tests 4.4-1 4.5 Safety Injection, Containment, Spray and Iodine Removal Systems Tests 4.5-1 4.6 Emergency Power System Periodic Tests 4.6-1 4.7 Main Steam Stop Valves 4.7-1 4.8 Auxiliary Feedwater System 4.8-1 4.9 Reactivity Anomalies 4.9-1 4.10 Environmental Radiation Survey 4.10-1 4.11 Refueling 4.11-1 4.12 Effluent Surveillance 4.12-1 4.13 Radioactive Material Source Leakage Test 4.13-1 4.14 Shock Suppressors (Snubbers) 4.14-1 4.15 Fire Suppression System Test 4.15-1 4.16 Overpressure Protection System 4.16-1 5.0 DESIGN FEATURES 5.1 Site 5.1-1 5.2 Containment Design Features 5.2-1 5.3 Reactor Design Features 5.3-1 5.4 Fuel Storage 5.4-1

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Proposed

TECHNICAL SPECIFICATIONS DEFINITIONS The following terms are defined for uniform interpreta-tion of the specifcations.

Thermal Power The rate that, the thermal energy generated by the fuel is accumulated by the coolant as it passes through the reactor vessel.

Reactor 0 eratin Modes Coolant .

Reactivity Temperature Mode ak oF k'10 Refueling T < 140 Cold Tavg

< 200 Shutdown Hot Tavg Shutdown

~ 580 Operating 0 Tavg

~f Any operation within the containment involving movement, of fuel and/or control rods when the vessel head is un-bolted.

~Oerable Capable of performing all intended functions in the intended manner.

1-1 PROPOSED

3.0 LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Coolant S stem to the operating status of the Reactor Coolant b'1'pplies System when fuel is in the reactor.

~b'o specify those conditions of the Reactor Coolant System which must be met, to assure safe reactor operation.

3.1.1 0 erational Com onents 3.1.1.1 Reactor Coolant Loo s a ~ When the reactor power is above 130 MWT (8.5%),

both reactor coolant loops and their associated steam generators and reactor coolant pumps shall be in operation.

b. If the conditions of 3.1.1.l.a are not met, then immediate power reduction shall be initiated under administrative control. If the shutdown margin/

meets the one loop requirements of Figure 3.10-2, then the power shall be reduced to less than 130 MWT. If the one loop shutdown margin of Figure 3.10-2 is not met, the plant shall be taken to the hot shutdown condition and the one loop shutdown margin shall be met.

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c. Except for special tests, when the average coolant temperature is above 350'F, or when the reactor is at hot shutdown or is critical with the reactor power less than or equal to 130 NWT (8.5%), at least one reactor coolant loop and its associated steam generator and reactor coolant pump shall be in opera-tion. The other loop and its associated must be operable so that. heat could- be steam'enerator removed via natural circulation. However, both reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2) core outlet temperature is maintained at least 10'F below.

saturation temperature.

d. If the conditions of 3.1.1.l.c are not met, suspend all operations involving a reduction in boron concentration of the Reactor Coolant: System and immediately initiate'orrective action to return the required coolant loop to operation.

e. When the reactor is at cold shutdown or when the average coolant temperature is between 200'F and 350'F, at least two of the following coolant loops shall be operable:

(i) reactor coolant loop A and its associated generator and reactor coolant pump.

~ steam IN&I ILL, II (ii) reactor coolant loop B and its associated steam generator and reactor coolant pump.

3.1-2 PROPOSED h

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(iii) residual heat removal loop.A.*

(iv) residual heat, removal loop B.*

While at cold shutdown or when the average coolant temperature is between 200'F and 350'F, at. least one of the coolant loops listed in paragraph 3.1.1.1.e shall be in operation. However, both reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> pro-vided 1) no operations are permitted that would cause dilution of ther reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10'F below saturation tempera-ture.

g. If the conditions of 3.1.1.1.e are not met, immediately initiate corrective action to return the required loops to operable status, and if not in cold shutdown already, be in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

If the conditions of 3.1.1.1.f are not met, then suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

• The normal or-;,emergency power source may be inoperable while in cold shutdown~

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3. 1-3 PROPOSED

i. At least one reactor coolant pump or the residual heat removal system shall be in operation when a reduction is made in the boron concentration of the reactor coolant.

j. At least one reactor coolant pump shall be in operation for a planned transition from one Reactor Operating Mode to another involving an increase in the boron concentration of the reactor coolant, except. for emergency boration.

k. A reactor coolant pump shall not. be started with one or more of the RCS cold leg temperatures

< 330'F unless 1) the pressurizer water volume is less than 324 cubic feet (38% level) or 2) the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.

3.1.1.2 Steam Generator

a. The temperature difference across the tube sheet,

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shall not exceed 100'F.

At least one pressurizer code safety valve shall be operable whenever the reactor head is bolted on the vessel.

b. Both pressurizer code safety valves shall be operable whenever the reactor is critical.

The plant is designed to operate with all reactor coolant loops in operation and maintain the DNBR above 1.30 during all normal 3.1-4 PROPOSED

operations and anticipated transients. Heat transfer analyses (1) show that reactor heat equivalent to 130 NWT (8.5%) can be removed by natural circulation alone. Therefore operation with one operating reactor coolant loop while below 130 NWT provides adequate margin.

The specification permits an orderly reduction in power if a reactor coolant pump is lost during operation between 130 MWT and 50% of rated power. Above 50% power, an automatic reactor trip will occur if either pump is lost. The power-to-flow ratio will be maintained equal to or less than one which ensures that the minimum DNB ratio increases at lower flow since the maximum enthalpy rise does not increase.

When the reactor coolant system average temperature is above 350'F, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require one loop be in operation and the other loop be capable of removing heat via natural circulation.

When the reactor coolant system average temperature is between 200'F and 350'F or while in cold shutdown, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be operable. Thus, if the reactor coolant loops are not operable, this specification requires two RHR loops to be operable.

When the borozf-'"co'ncenCration of the reactor coolant system is to be reduced the process must be uniform to prevent sudden reactivity 3.1-4a PROPOSED

changes in the reactor. Mixing of the reactor coolant will be sufficient to prevent a sudden increase in reactivity if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the primary system volume in approximately one half hour. The pressurizer is of no concern because of the low pressurizer volume and because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant. When the boron concentration of the reactor coolant system is to be increased, the process must be uniform to prevent sudden reactivity increases in the reactor during subsequent startup of the reactor coolant pumps. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump is running while the change is taking place. Emergency boration without a reactor coolant pump in operation is not prohibited by this specification.

Prohibiting reactor coolant pump starts without a large void in the pressurizer or without a limited RCS temperature differential will prevent RCS overpressurization due to expansion of cooler RCS water as it enters a warmer steam generator. A 38% level in the pressurizer will accommodate the swell resulting from a reactor coolant pump start with a RCS temperature of 140'F and steam generator secondary side temperature of 340'F, or the maximum temperature which usually exists prior to cooling the reactor with-"-CKe"-RHR"'system.

3.1-4b

Temperature requirements for the steam generator correspond with measured NDT for the shell and allowable thermal stresses in the tube sheet.

Each of the pressurizer code safety valves is designed to relieve 288,000 lbs. per hr. of saturated steam at the valve set point.

Below 3SO'F and 350 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available the amount of steam which could be generated at, safety valve relief pressure would be less than half the valves'apacity. One valve therefore provides adequate defense against overpressurization.

References (1) FSAR Section 14.1.6 (2) FSAR Section 7.2.3

3. 1-4c

least, one source range netron flux monitor shall be in service.

d. At least one residual heat removal loop shall be in operation.*

e. Immediately before reactor vessel head removal and while loading and unloading fuel from the reactor, the minimum boron concentration of 2000 ppm shall be maintained in the primary coolant system and checked by sampling twice each shift.

Direct communication between the control room and the refueling cavity manipulator crane shall be available whenever changes in core geometry are taking place.

g. In addition to the reguirements of paragraph 3.8.l.d, while in the refueling mode with less than 23 feet of water above the top of the reactor vessel flange, two residual heat removal loops shall be operable.*

h. During movement of fuel or control rods within the reactor vessel cavity, at least 23 feet. of water shall be maintained over the top of the reactor vessel

• Either the normal or the emergency power source may be inoperable for each residual. heat;.removal loop.

C'~OS '~qeew wee e 3.8-2 PROPOSED

flange. If this condition is not met, all operations involving movement of fuel or control rods in the reactor vessel shall be suspended.

3.8.2 If any of the specified limiting conditions for refueling is not met., refueling of the reactor shall cease; work shall be initiated to correct the violated conditions so that the specified limits are met; no operations which may increase the reactivity of the core shall be made.

Basis:

The equipment. and general procedures to be utilized during refueling are discussed in the FSAR. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that. no incident could occur during the refueling operations that would result in a hazard

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3. 8-2a PROPOSED

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provided on the lifting hoist to prevent movement of more than one fuel assembly at a time. The spent fuel transfer mechanism can accommodate only one fuel assembly at a time. In addition interlocks on the auxiliary building crane will prevent the trolley from being moved over storage racks containing spent fuel.

References:

(1) FSAR Section 9.5.2 (2) Table 3.2.1-1 (3) FSAR - Section 9.3.1

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Amendment 11 3. 8-4 PROPOSED

e. Charcoal adsorbers shall be installed in the venti-lation system exhaust from the spent fuel storage pit area and shall be operable.

3. 11.2 Radiation levels in the spent fuel storage area shall be monii ored continuously.

3.11.3 The trolley of the auxiliary building crane shall never be stationed or permitted to pass over storage racks containing spent fuel.

3.11.4 Fuel assemblies with less than 60 days since irradiation shall not be placed in storage positions with less spacing between them than that indicated in Figure 3.11-1 by the designation RDF.

3.11.5 The spent fuel pool temperature shall be limited to 150'F.

3.11.6 The spent. fuel shipping cask shall not be carried by the

'auxiliary building crane, pending the evaluation of the spent, fuel cask drop accident and the crane design by RGGE and NRC review and approval.

Basis:

Charcoal adsorbers will reduce significantly the consequences of a refueling accident which considers the clad failure of a single irrad-iated fuel assembly. Therefore, charcoal adsorbers should be employed whenever irradiated fuel is being handled. This requires that the ventilation system should be operating and drawing air through the adsorbers.

The desired.='air.-..<low=-.path, ~

when handling irradiated fuel, is from the outerde oPt'he uiidrng into the operating floor areatowa,r d the spe'nt'uel'"storage pit, into the area exhaust ducts, through the Change 4 3.11-2 PROPOSED Amendment 11, 34

adsorbers, and out through the ventilation system exhaust to the facility vent. Operation of a main auxiliary building exhaust fan assures that air discharged into the main ventilation system exhaust duct will go through a HEPA and be discharged to the facility vent.

Operation of the exhaust fan for the spent fuel storage pit area causes air movement on the operating floor to be towards the pit.

Proper operation of the fans and setting of dampers would result in a negative pressure on the operating floor which will cause air leakage to be into the building. Thus, the overall air flow is from the location of low activity (outside the building) to the area of highest activity (spent fuel storage pit). The exhaust air flow would be through a roughing filter and charcoal before being discharged from the facility. The roughing filter protects the If adsorber from becoming fouled with dirt; the adsorber removes iodine, the isotope of highest radiological significance, resulting from a fuel handling accident. The effectiveness of charcoal for removing iodine is assured by having a high throughput and a high removal efficiency. The throughput is attained by operation of the exhaust fans. The high removal efficiency is attained by minimizing the amount of iodine that bypasses the charcoal and having charcoal with a high potential for removing the iodine that, does pass through the charcoal.

The minimum spacing specified for fuel assemblies with less than 60 days decay is based on maintaining the potential release of fission proddcCs= that"could occur should an object fall on and dam~a e stox'ed fuel to less than that which could have occurred with fuel 'stored in the original fuel'torage racks.

3 11 3

The spent, fuel pool temperature is limited to 150'F because if the spent fuel pool cooling system is lost at that temperature, suf-ficient time (approximately 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />) is available to provide back-up cooling, assuming the maximum anticipated heat load (full core discharge 6 previously stored fuel), until a temperature of 180'F is reached, the temperature at which the structural integrity of the pool was analyzed and found acceptable.

References (1) FSAR Section 9.3-1 (2) ANS-5.1 (N 18.6), October 1973

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Proposed

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4.3 REACTOR COOLANT SYSTEM

~11'pplies to surveillance of the reactor coolant system and its components.

To ensure operability of the reactor coolant system and its components.

S ecifications:

4.3.1 Reactor Vessel Material Surveillance Testing 4.3.1.1 The reactor vessel material surveillance testing program is designed to meet the requirements of Appendix H to 10 CFR Part 50. This program consists of the, metal-lurgical specimens receiving the following test: tensile, charpy impact and the WOL test. These tests of the Radiation Capsule Specimens shall be performed as follows:

~Ca sule Time Tested End of 1st core cycle End of 3rd core cycle 10 years, at, nearest refueling 20 years, at nearest refueling 30 years, at, nearest refueling N Standby

'.3.1.2 The report of the Reactor Vessel Material Surveillance shall be written Q~~LCIAOl" as a Summary Technical Report as required

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i eac4o~Coolant Loops 4.3-1 Proposed

4.3.2.1 When reactor power is above 130 NWt (8.5%), the reactor coolant pumps shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.3.2.2 When the average coolant temperature is above 350'F but the reactor is not critical, when the reactor is at hot shutdown, or when the reactor is critical but'reactor power is less than or equal to 130 NWt,(8.5%):

a) the operating reactor coolant pump(s) shall be verified to be in operation and circulating reactor coolant at, least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> b) reactor coolant pump which is not operating,, but must be operable, shall be demonstrated operable once per 7 days by verifying correct breaker alignments and indicated power availability.

4.3.2.3 When the reactor is at cold shutdown or when the average coolant temperature is between 200'F and 350'F, and fuel is in the reactor, the following shall be performed to demonstrate a loop is operable. Tests need not be per-,

formed if a loop is not relied upon to satisfy the re-quirements of Specification 3.1.1.1.e.

a) to demonstrate a reactor coolant loop operable, the reactor coolant pump(s), if not in operation, shall be demonstrated operable at least once per 7 days by verifying correct breaker alignments

., ..~~~and indicated. power availability.

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4.3-2 Proposed

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b) to demonstrate a residual heat removal pump is operable, the surveillance specified in the Inservice Pump and Valve Test Program prepared pursuant to 10 CFR 50.55a shall be performed.

4.3.2.4 When the reactor is at cold shutdown or when the average coolant temperature is between 200'F and 350'F and fuel is in the reactor at least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Basis:

This material surveillance program monitors changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region of the reactor resulting from exposure to neutron irradiation and the thermal environment. The test. data obtained from this program will be used to determine the conditions under which the reactor vessel can be operated with adequate margins of safety against fracture throughout its service life.

4.3-3 Proposed

4.11 Refuelin

~11'pplies to refueling and to fuel handling in the spent.

fuel pit.

Fuel:Pit Charcoal Adsorber System

'f'.11.1 Spent 4.11.1.1 Within 60 days prior to each major fuel handling, the spent fuel pit charcoal adsorber system shall have the following conditions demonstrated. After the conditions have been demonstrated, the occurrence of painting, fire or chemical release in any ventilation zone communicat-ing with the spent fuel pit charcoal adsorber system shall require that the following conditions be redemonstrated before major fuel handling* may continue.

a. The total air flow rate from the charcoal adsorbers shall be at least 75% of that measured with a com-piete set of new adsorbers.

b. In-place Freon testing, under ambient. conditions, shall show at least 99% removal.

c. The results of laboratory analysis on a carbon sample shall show 90% or greater radioactive methyl iodide removal when tested at at least 150'F and 95% RH and at 1.5 to 2.0 mg/m loading

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d. Flow shall be maintained through the system using either the filter or bypass flow path for at least 15 minutes each month.

4.11.1.2 After each replacement of a charcoal filter drawer or after any structural maintenance on the charcoal housing for the spent fuel pit charcoal adsorber system, the condition of Specification 4.11. 1. 1.b shall be demonstrated for the affected portion of the system.

4.11.2 Residual Heat. Removal and Coolant Circulation 4.11.2.1 When the reactor is in the refueling mode and fuel is in the reactor, at least one residual heat, removal loop shall be verified to be in operation and circulating reactor coolant at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.11.2.2 When the water level above the top of reactor vessel flange is less than 23 feet, both RHR pumps shall be verified to be operable by performing the surveillance specified in the Inservice Pump and Valve Test Program prepared pursuant to 10 CFR 50.55a.

4.11.3 Water fevel Reactor Vessel 4.11.3.1 The water level in the reactor cavity shall be determined to be at least its minimum reguired depth within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to the start of and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during movement of fuel assemblies or control rods in containment.

Basis emeasuremen assures that air is being withdrawn Xrom.'the:;sjent-. Xiiej..-.pig area,.and passed through the adsorbers. The flow is measured prior to employing the adsorbers to establish that 4.11-2

there has been no gross change in performance since the system was last used. The Freon test provides a measure of the amount of leakage from around the charcoal adsorbent.

The ability of charcoal to adsorb iodine can deteriorate as the charcoal ages and weathers. Testing the capacity of the charcoal to adsorb iodine assures that an acceptable removal efficiency under operating conditions would be obtained. The difference between the test, requirement of a removal efficiency of 90% for methyl iodine and the percentage assumed in the evaluation of the fuel handling accident provides adequate safety margin for degrada-tion of the filter after the tests.

Retesting of the spent fuel pit charcoal adsorber system in the event, of painting, fire, or chemical release is required only if the system is operating and is providing filtration for the area in which the painting, fire, or chemical release occurs.

Testing of the air filtration systems will be tested, to the extent: it can be given the configuration of the systems, in accordance with ANSI N510-1975, "Testing of Nuclear Air-Cleaning Systems".

Reference:

(1) Letter from E. J. Nelson, Rochester Gas and Electric Corporation to Dr. Peter A. Morris, U.S. Atomic Energy Commission, dated February 3, 1971 CCQskj i ~ i 4.11-3

Attachment B Technical Specifications have been prepared for decay heat removal system which generally follows the guidance of the letter from Darrel G. Eisenhut dated June 11, 1980. Since a new Technical Specification addressing water level during refueling operation is being proposed, NRC guidance as provided by Thomas M. Novak's letter dated August 15, 1980 is also being incorporated.

The conditions for one loop operation have not been revised.

One loop operation has recently been reviewed by the NRC under the Systematic Evaluation Program and found to be acceptable (see the letter dated May 29, 1979 from Dennis L. Ziemann).

Analyses have shown that natural circulation can remove heat, up to the equivalent of 130 MWt (8.5%). Thus, the single failure of the operating reactor coolant pump while below 130 MWt (8.5%)

does not. result in an unacceptable condition.

At low temperature conditions, two loops are required to be operable. This ensures that, in the event of a single failure causing loss of one loop, residual heat can still be removed.

During refueling when the reactor cavity is filled, only one loop need be operable. This is due to the substantial heat sink .

offered by the water in the cavity.

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