Text

Proposed TS Section 1.6.1 Re Power in Any Core Quadrant & 1.6 Re Quadrant Power Tilt
	ML20044B693
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Site:	Arkansas Nuclear
Issue date:	02/24/1993
From:	; ENTERGY OPERATIONS, INC.
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{{#Wiki_filter:_._ _ F b ATTACHMENT l

PROPOSED TECHNICAL SPECIFICATION AND 2 RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING , I.ICENSE NO. DPR-51 i ENTERGY OPERATIONS, INC. ARKANSAS NUCLEAR ONE, UNIT ONE . DOCKET NO. 50-313 l i i 1 I

i 1 l 9303030075 930224 PDR .ADOCK 05000313 P PDR l + l

i , Description of Changes to ANO-1 Technical Specifications LOCATION DESCRIPTION BACKGROUND / DISCUSSION l Section 1.6.1 Change reference from: This was a typographical error in the original Page 4 " specification 3.5.24" to: specification. There is no specification 3.5.24.

                                                                                     " specification 3.5.2.4"                               The correct reference is specification 3.5.2.4.

Section 1.6.1 Change equation from: This change removes ambiguity introduced in the Page 4 original specification and does not reflect a Power in any core quadrant 1 change in the power imbalance calculation Average power of all quadrants method. The revised equation format corresponds j

to that in Topical Report BAW-10122A as to..

referenced in section 3A.8 of ANO-1 Safety e . Analysis Report, and accepted by NRC, ww in any cme qua& ant _; 3 September 14,1979. 100 ( Average power of all quadrants j Bases of Section 3.0.5 Change "3.7.2.C provides for a 2 day Submittal mark-up for amendment 57 appears to Page 15b-3 out-of-service time" to "3.7.2.C provides have been mistyped. Specification 3.7.2.C allows for a 7 day out-of-service time" for a 7 day out-of-service period for either of the two diesel generators. Section 3.1.1 Change " operable" to " inoperable" and Submittal mark-up for amendment 57 appears to Page 16 capitalize " Hot Shutdown" in section have been mistyped. This specification identifies 3.1.1.3.A. Add degree symbol to actions to be taken when one pressurizer code temperature specification in 3.1.1.2.A. safety valve is inoperable. As written it would require the operable valve to be restored to i operable status rather than the inoperable one. For consistency, the defined term " Hot Shutdown", added in amendment 57, is being capitalized.

Description of Changes to ANO-1 Technical Specifications (continued) LOCATION DESCRIPTION BACKGROUND / DISCUSSION Section 3.1.1.5 Change " reduced" to " reduce" This action statement requires operators to restore Page 16a an inoperable coolant loop with in 72 hours or reduce coolant Tave to 280*F in 12 hours. This typographical error was introduced in amendment 56. Section 3.1.5 in sections 3.1.5.1, 3.1.5.2 and 3.1.5.3, These errors were present in the original Limiting Page 25 add degree symbols to temperature Conditions for Reactor Coolant chemistry. The values. In section 3.1.5.4 insert comma proposed changes are for clarity and consistency. in first sentence following " exceed 1.0 ppm." At the end of the first sentence of the Bases, move the footnote symbols to the end of the preceding line and change them to superscript. Section 3.3.1 Change " low pressure injection" to " Low This specification sets operability requirements for Page 36 Pressure injection (LPI)" in section the emergency core cooling and reactor building 3.3.1.D. in section 3.3.1.F insert spray systems. These changes are proposed to

                                                                  " Borated Water Storage Tank" before                                               provide clarity and consistency in references to acronym "BWST."                                                                    major plant systems. These errors were present in the original specification.

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a,.~-- ,,.-n-.-- w. . - - , - - . .-. .- - - - n,---, ,- - - - - -- --- ,,,. -,,-, , . ., ~ . , , _ . . , - - - , -,

Description of Changes to ANO-1 Technical Specifications (continued) LOCATION DESCRIPTION BACKGROUND / DISCUSSION Section 3.4.4 & 3.4.5 Change all occurrences of " specification" This specification establishes the Limiting Page 40a to " specifications"in section 3.4.4. Condition for Operation for the capability to Capitalize each occurrence of " hot remove decay heat from the reactor core. The l shutdown" and " cold shutdown" in operational conditions in specification 3.4.5 are section 3.4.5. being capitalized to indicate that these are defined terms. This error was introduced in amendment

50. In specification 3.4.4, the word " specification" is used three times in reference to more than one section. Each occurrence of" specification"is being replaced with " specifications" to correct this typographical error introduced in amendment 91.

Bases Delete comma following "for protective Specification 3.5.1 applies to instrumentation and Section 3.5.1 action"in tenth paragraph. Insert circuitry necessary to assure reactor safety. Page 43 hyphens into logic descriptors "two out of Hyphens are being inserted into the logic four", "two out of three", and "one out of descriptors for clarification. The comma is being two"in third paragraph. removed to correct a grammatical error. Both of these errors were present in the original specification. Bases Change footnote symbols to superscript. Section 3.5.2 applies to Control Rod Groups and Section 3.5.2 Capitalize " group" and " groups" in last Power Distribution Limits. The footnote symbols Page 48a paragraph. Delete " Babcock & Wilcox" are being changed to superscript and the , from lower-right corner. references to control rod " group 1" and " groups 6 and 7" are being capitalized for consistency. The "B&W" logo was inadvertently copied from the B&W Standard Technical Specifications and is being deleted. These errors were incorporated in amendment 52.

 ~__ _ _ _ ___ ____ _ -__ - .                         . - ,       - . - . . - . - .      _ -. -       - ._     .      .-              ._.   .. -      - -    _ _ _     . . . _ . _ _ .

t a Description of Changes to ANO-1 Technical Specifications (continued) LOCATION DESCRIPTION BACKGROUND I DISCUSSION Table 4.1-1 Change "SGS or B" to "SGA or B" Typographical error in request for amendment 135. Page 72b1 There is no "SGS" associated with ANO-1. Section 4.4.1.1 Change " exercised" to " exercise" in These errors were in the original specification. Page 80 specification 4.4.1.1.3.d. Underline Titles They are being corrected for consistency and of TS 4.4.1.1.3 and 4.4.1.1.4 and change grammatical accuracy.

                                                   " shut down" to " shutdown" in specification 4.4.1.1.4.

Section 4.4.2.1.2 Change " extent and cause change" to Grammatical error in original specification. Page 85a " extent and cause of change" Section 4.8.1.a.1 Insert "of 5 minutes, and devebps a These words were inadvertently deleted while Page 105 discharge pressure of " following preparing the submittal for amendment 121.

                                                   " operates for a minimum".

Section 4.30.1.2.b Change " plan" to " plant". Typographical error in request for amendment 118. Page 110ss This misspelling occurred in a reference to plant effluents. Section 4.30.1.1 Change Table reference "4.30.2" to These inconsistencies were introduced in Page 110ss "4.30-2". Change reference "shown in amendment 118. table 4-1" to " listed in Table 4-1" Section 4.30.1 Align " Applicability" and " Objective" These format errors were introduced in Page 110ss sections with margin, delete colons from amendment 88 which inserted page 110ss into the titles, and change section Title ANO-1 Technical Specifications

                                                   " Specifications" to " Specification".

Description of Changes to ANO-1 Technical Specifications (continued) LOCATION DESCRIPTION BACKGROUND / DISCUSSION Section 4.30.1.2.c Change " Table 4.29-1" to " Table 4.30-1" Typographical error in request for amendment 88. Page 110tt Table 4.29-1 does not contain information relevant to section 4.30.1.2.c

Section 5.2 Change reference 2 from in the original specification, footnote 2 to l Page 113 "FSAR Section 5.1.5" to "FSAR Section specification 5.2.2 " Reactor Building isolation 5.2.5" and change footnote symbols to System" incorrectly referenced FSAR Section superscript. 5.1.5, " Wind and Tomado Loads" rather than Section 5.2.5, " Isolation System." The format of the original footnote symbols is being revised for consistency.

Section 5.4.1.1 Change "Keff of less than .9" to This change adds a leading zero for consistency. Page 116 "Kerg of less than 0.9" The zero was omitted in the original specification. t Section 6.5.1.6 Insert "and" following "in intent thereto," This specification, concerning the responsibilities Page 121 in section 6.5.1.6.a. Underline title of the Plant Safety Committee (PSC), is being

                                                                        " Responsibilities".                                       changed to remove ambiguity and correct format.

Section 6.12.2.1 Insert colon following " testing shall be This specification addresses the reporting Page 140 submitted following". Remove open requirements of Title 10 of the Code of Federal parentheses "(" preceding list numbers. Regulations. The proposed change is intended to correct the format and clarify the specification.

b Y

An evaiustion of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(c). A discussion of those standards in 10CFR50.92(c)is discussed below.

Criterion 1 Does not involve a significant increase in the probability or consequence of an accident previously evaluated. I These changes do not affect the intent of any specification. Also, the proposed changes do not provide any relief from the requirements of the TS, or change the intended operation or administrative requirements of the plant orits design basis. The proposed changes clarify the existing specification requirements and are ' administrative in nature. Since they are administrative in nature, these changes do not significantly increase the ; probability or consequence of any previously analyzed accident occurring . ' Criterion 2 Does not create the possibility of a new or different kind of accident from any previously evaluated. . J > The proposed changes do not involve any design changes, plant modifications or y changes in plant operation; rather, they only reflect a more accurate description of the , specification requirements. 1 i The proposed changes clarify the existing specification requirements and are ,

administrative in nature.

J 1 Since they are administrative in nature, these changes do not create the possibility of a j new or different kind of accident from any previously evaluated. l t Criterion 3 Does not involve a significant reduction in a margin of safety. , l i The proposed changes only clarify the existing requirements. They do not relax any j i specification requirements. ! i The proposed changes are administrative in nature and do not affect any plant safety , 4 parameters, accident mitigation capabilities, or margin of safety. I 1t ] Since these changes are administrative in nature, they do not involve a significant reduction in a margin of safety. The commission has provided guidance in 51 FR 7750 dated March 6,1986, conceming the application of these 10CFR50.92 standards by providing examples of amendments which are likely to involve no significant hazards considerations. The proposed amenoment most closely matches example (i): "A purely administrative change to technical specifications: for example, a change to achieve consistency i throughout the technical specifications, correction of an error, or a change in ; nomenclature." ! - i 1

I f Therefore, based on the reasoning presented above and the previous discussion of the [ amendment request Entergy Operations has determined that the requested changes l do not involve a significant hazards consideration. > l I

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sa Y 4 ' 4 - e > f l i I i i I i F PROPOSED TECHNICAL SPECIFICATION CHANGES h i i n t I r 5 I i 9 [ t I t 5 i I 4 l l l l 1 i i y

         .                                                                                     f 1.5.4       Instrument Channel Calibration                                     ;

An instrument channel calibration is a test, and adjustment (if l necessary), to establish' that the channel output responds with acceptable range and accuracy to.known values.of the parameter which the : channel measures or an accurate simulation of these values. Calibration shall encompass the entire channel, including equipment actuation, alarm i or trip and shall be deemed to include the channel test. 1.5.5 Heat Balance Check A heat balance check is a comparison of the indicated neutron power and ; core thermal power. ( 1.5.6 Heat Balance Calibration s An adjustment of the power range channel amplifiers output to agree with j the core thermal power as determined by a weighted primary and secondary i i heat balance considering all heat losses. Between 0 and 15% power, only l l the primary heat balance is considered. From 15 to 100% power the heat ,

!               balance is weighted linearly with only the secondary heat balance being        (

considered at 100% power. l 1.6 POWER DISTRIBUTION j t 1.6.1 Ouadrant Power Tilt j 0 Quadrant power tilt is defined by the following equation and is

expressed in percent i [ Power in any core quadrant \

j 100 -1 l ,

( Average power of all quadrants { f The power in any quadrant is determined from the power range channel .l displayed on the console for that quadrant. The average power is ; determined from an average of the outputs of the power range channels. ! If one of the power range channels is out' of service, the remaining l three operable power range channels or the incore detectors will be used to determine the average power. The quadrant power tilt limits as a : function of power are stated in Specification 3.5.2.4. l ! f 1.6.2 Reactor Power Imbalance Reactor power imbalance is the power in the top half of the'cora minus ; the power in the bottom half of the' core expressed as a percentage of [ rated power. Imbalance is monitored continuously by the RPS using input ! from the power range channels. Imbalance limits are defined in Specification 2.1 and imbalance setpoints are defined in Specification 2.3. ) I i 4 ,

4 BASES (continued) initiated or that higher modes of operation are not entered when corrective action is being taken to obtain compliance with a Specification by restoring equipment to OPERABLE status or parameters to specified limits. Compliance with Action requirements that permit continued operation of the facility for an unlimited period of time provides an acceptable level of safety for continued operation without regard to the status of the plant before or after a mode change. Therefore, in this case, if the requirements. for continued operation have been met in accordance with the requirements of the specification, then entry into that mode of operation is permissible. The provisions of this specification should not, however, be interpreted as endorsing the failure to exercise good practice _ in restoring systems or components to OPERABLE status before plant startup. When a shutdown is required to comply with Action requirements, the provisions of Specification 3.0.4 do not apply because they would delay placing the facility in a lower mode of operation. For the purpose of compliance with this specification the term ' shutdown' is defined as a required reduction in the REACTOR OPERATING CONDITION. 3.0.5 Delineates what additional conditions must be satisfied to permit operation to continue when a normal or emergency power source is not OPERABLE. It specifically prohibits operation when one division is inoperable because its normal or emergency power source is inoperable and a system, subsystem, l train, component or device in another division is inoperable for another reason. The provisions of this specification permit the Limiting Condition for Operation statements associated with individual systems, subsystems, trains, components or devices to be consistent with the Limiting Condition for Operation statements of the associated electrical power source. It allows operation to be governed by the time limits of the Limiting Condition for Operation for the normal or emergency power source, not the individual Limiting Condition for Operation statements for each system, subsystem, train, component or device that is determined to be inoperable solely because of the inoperability of its normal or emergency power source. For example, Specification 3.7.2.C provides for a 7 day out-of-service time l when one emergency diesel generator is not OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the-inoperable emergency power source would also be inoperable. This would dictate invoking the applicable Action statements for each of the applicable Limiting Conditions for Operation. However, the provisions of Specification 3.0.5 permit the time limits for continued operation to Amendment No. 57, 161 15b-3

3.1 REACTOR COOLANT SYSTEM Anolicability Applies to tne operating status of the reactor coolant system. Obiective Tc specify those limiting conditions for operation of the reactor coolant system which must be met to ensure safe reactor operations. 3.1.1 Operational Components Soecification 3.1.1.1 Reactor Coolant Pumps A. Pump combinations permissible for given power levels shall be as shown in Table 2.3-1. B. The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant. With no reactor coolant pumps or decay heat removal pumps running, immediately suspend all operations involving a reduction of boron concentration in the reactor coolant system. 3.1.1.2 Steam Generator A. Two steam generators shall be operable whenever the reactor coolant average temperature is above 280F. 3.1.1.3 Pressurizer Safety Valves A. Both pressurizer code safety valves shall be operable when the reactor is critical. With one pressurizer code safety valve inoperable, either restore the valve to operable status within 15 minutes or be in HOT SHUTDOWN within 12 hours. B. When the reactor is subcritical, at least one pressurizer code safety valve shall be operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code, Section III. Tha provisions of Specification 3.0.3 are not applicable. 3.1.1.4 Reactor Internals Vent Valves The structural integrity and operability of the reactor internals vent valves shall be maintained at a level consistent with the acceptance criteria in Specification 4.1. The provisions of Specification 3.0.3 are not applicable. 3.1.1.5 Reactor Coolant Loops A. With the reactor coolant average temperature above 280 F, the reactor coolant loops listed below shall be operable: Amendment No. 27,5%, 57 16

i s I 1

1. Reactor Coolant Loop (A) and at least one associated !

reactor coolant pump.

2. Reactor Coolant Loop (B) and at Icast one associated !

reactor coolant pump. , Otherwise, restore the required loops to operable status within 72 hours or reduce the reactor coolant average .l temperature to less than or equal to 280'F within the next 12 hours. B. With the reactor coolant average temperature above 280*F, I at least one of the reactor coolant loops listed above j shall be in operation. i Otherwise, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and . immediately initiate corrective action to return the required loop to operation. 3.1.1.6 Decay Heat Removal !

                                                                                                                 ),
                                                                                               ~

With the reactor coolant average temperature at or below 8 280'F, but the reactor above the refueling shutdown condition, ; at least two of the coolant loops listed below shall be { operable, and at least one loop shall be in operation:* j ' i.

1. Reactor Coolant Loop (A) and its associated steam :

> generator and at least one assoc.tated reactor coolant ; pump.

2. Reactor Coolant Loop (B) and its associated steam j generator and at least one associated reactor coolant ;

Pump. ; P

3. Decay Heat Removal Loop (A)** j t

4. Decay Heat Removal Loop (B)** j

                                                                                      .                           i A. With less than the above required coolant loops OPERABLE,.              j immediately initiate corrective action to return the required            l coolant loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours.                                           .
                                                                                                                  ?

I B. With no coolant loop in operation, 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. l 1

                       *All reactor coolant pumps and decay heat removal pumps may be                             ,

de-energized for up to I hour provided (1) no operations are permitted j 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. , i 1

                       **The normal or emergency power source may be inoperable when the                          ,

reactor is in a cold shutdown condition. l Amendment No. 56 16a j i

             . - - .        - - - . . -            ,-       -    - - - -         - ..      ..    - -- -.   -    .h

3.1.5 Chemistry Acolicability Applies to the limiting conditions of reactor coolant chemistry for continuous operation of the reactor. Abiective To protect the reactor coolant system from the effects of impurities in the reactor coolant. Snecification 3.1.5.1 The following limits shall not be exceeded for the listed reactor coolant conditions. Contaminant Snecification Peactor Coolant Conditions Oxygen as 0 2 0.10 ppm max above 250'F l Chloride as Cl- 0.15 ppm max above cold shutdown conditions Fluoride as F~ 0.15 ppm max above cold shutdown conditions 3.1.5.2 During operation above 250'F, if any of the l specifications in 3.1.5.1 is exceeded, corrective action shall be initiated within 8 hours. If the concentration limit is not restored within 24 hours after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using normal procedures. 3.1.5.3 During operations between 250'F and cold shutdown ! conditions, if the chloride or fluoride specification in - 3.1.5.1 are exceeded. corrective action shall be initiated within 8 hours to restore the normal operating limits. If the specifications are not restored within 24 hours after initiation of corrective action, the reactor shall be placed in a cold shutdown condition using normal procedures. 3.1.5.4 If the oxygen concentration and either the chloride or fluoride concentration of the primary coolant system exceed 1.0 ppm, the reactor shall be immediately brought to the hot l shutdown condition using normal shutdown procedures, and action is to be taken immediately to return the system to within normal operation specifications. If specifications given in 3.1.5.1 have not been reached in 12 hours, the , reactor shall be brought to a cold shutdown condition using I normal procedures. Bases By maintaining the chloride, fluoride, and oxygen concentration in the reactor coolant within the specifications, the integrity of the reactor coolant system is protected against potential stress corrosion attack (2,2). I 25

4 3.3 -EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING AND REACTOR BUILDING SPRAY SYSTEMS Aonlicability Applies to the emergency core cooling, reactor building emergency cooling and reactor building spray systems. Obiectivity To define the conditions necessary to assure immediate availability of the emergency core cooling, reactor building emergency cooling and reactor building spray systems. i Soecification 3.3.1 The following equipment shall be operable whenever containment f integrity is established as required by Specification 3.6.1: l (A) One reactor building spray pump and its associated spray nozzle header. (B) One train of reactor building emergency cooling. (C) Two out of three service water pumps shall be operable, powered from independent essential buses, to provide redundant and independent flow paths. (D) Two engineered safety feature actuated Low Pressure Injection (LPI) pumps shall be operable. l (E) Both low pressure injection coolers and their cooling water supplies shall be operable. (F) Two Borated Water Storage Tank (BWST) level instrument channels l I shall be operable. (G) The borated water storage tank shall contain a level of 40.2 1 1.8 _ft. (387,400 1 17,300 gallons) of water having a concentration of 2470 200 ppm boron.at a temperature not less than 40F. The manual valve on the discharge , line from the borated water storage tank shall be locked j open. (H) The four reactor building- emergency sump isolation valves to the LPI system shall be either manually or remote-manually operable. Amendment No. JS,JS ,JJJ ,Jf p,145 36

3.4.3 The automatic steam generator isolation system within EFIC shall be operaule when main steam pressure is greater than 750 psig. L.4.4 Components required to be operable by Specifications 3.4.1, l 3.4.2, and 3.4.3 shall not be removed from service for more than 24 consecutive hours. If the system is not restored to meet the requirements of Specifications 3.4.1, 3.4.2 and 3.4.3 within 24 hours, the reactor shall be placed in the hot shutdown condition within 12 hours. If the requirements of Specifications 3.4.1, 3.4.2, and 3.4.3 are not met within an additional 48 hours, the reactor shall be placed in the cold shutdown condition within 24 hours. , 3.4.5 If the condition specified in 3.4.1.4 cannot be met: 1 1. With one EFW flow path inoperable, the unit shall be brought ! to HOT SHUTDOWN within 36 hours, and if not restored to an l l operable status within the next 36 hours, the unit shall be ; brought to COLD SHUTDOWN within the next 12 hours or at the l l maximum safe rate. l 4 t

2. If both EFW trains are inoperable, restore one train to f
operable status within one hour or be in HOT SHUTDOWN within the next 6 hours and COLD SHUTDOWN within the next 12 hours or j

' at the maximum safe rate. l

3. If both EFW trains and the AFW pump are inoperable, the unit i shall be immediately run back to 55% full power with feedwater j j supplied from the MFW pumps. As soon as an EFW train or the i AFW train is operable, the unit shall be placed in COLD i SHUTDOWN within the next 12 hours or at the maximum safe rate. [

i ! 1 ! l t i 3 ! I j' s i i i f l e i l Amendment No. 55, 91 40a t _ __ _ -- - , - ,_ , i

Bases Every reasonable effort will be made to maintain all safety instrumentation in operation. A startup is not permitted unless the requirements of Table 3.5.1-1, Columns 3 and 4, are met. Operation at rated power is permitted as long as the systems have at least the redundancy requirements of Column 4 (Table 3.5.1-1). This is in agreement with redundancy and single failure criteria of IEEE 279 as described in FSAR, Section 7. There are four reactor protection channels. Normal trip logic is two-out-of-four. Required trip logic for the power range instrumentation channels is two-out-of-three. Minimum trip logic on other instrumentation channels is one-out-of-two. The four reactor protection channels were provided with key operated bypass switches to allow on-line testing or maintenance on only.one channel at a time during power cperation. Each channel is provided with alarm and lights to indicate when that channel is bypassed. There will be one reactor protection system channel bypass switch key permitted in the control room. Each reactor protection channel key operated shutdown bypass switch is provided with alarm and lights to indicate when the shutdown bypass switch is being used. The source range and intermediate range nuclear flux instrumentation scales overlap by one decade. This decade overlap will be cchieved at 10-28 amps on the intermediate range acale. The ESAS employs three independent and identical analog-channels, which supply trip signals to two independent, identical digital subsystems. In order to actuate the safeguards systems, two out of three analog channels must trip. This will cause both algital subsystems to trip. l Tripping of either digital subsystem will actuate all safeguards systems associated with that digital subsystem. Because only one digital subsystem is necessary to actuate the safeguards systems and these systems are capable of tripping even when they are being tested, a single failure in a digital subsystem cannot prevent protective action. Removal of a module required for protection from a RPS channel will cause that channel to trip, unless that channel has been bypassed, so that only one channel of the other three must trip to cause a reactor trip. Thus, sufficient redundancy has been built into the system to cover this situation. Removal of a module required for protective action from an analog ESAS l channel will cause that channel to trip, so that only one of the other two must trip to actuate the safeguards systems. Removal of a module required

43

                                               ~                .       .

I simultaneously all other engineering and uncertainty factors are also at their limits.* Conservatism is introduced by application of: +

a. Nuclear unce.t,saty factors. -f

b. Thermal calibration.

c. Fuel densification effects.

d. Hot rod manufacturing tolerance factors. I

e. Fuel rod bowing. ;

The 20 15 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower parts of the stroke. Control rods are arrange! in groups or banks defined as follows: Group Function 1 Safety 2 Safety 3 Safety ; i 4 Safety 5 Regulating 6 Regulating i 7 Regulating 8 APSR (axial power shaping bank) l The rod position limits are based on the most limiting of the following ; three criteria: ECCS power peaking, shutdown margin, and potential' ejected red worth. As discussed above, compliance with the ECCS power - peaking criterion is ensured by the rod position limits. The minimum , - available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the , highest worth control rod that is withdrawn remains in the full-out , position (2). The rod position limits also ensure that inserted rod l groups will not contain single rod worths greater than 0.65% Ak/k at i rated power. These values have been shown to be safe by the safety ! analysis (8) of the hypothetical rod ejection accident. A maximum l single inserted control rod worth of 1.0% Ak/k is allcwed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0% Ak/k at beginning of life, hot zero power, would result in a lower transient peak thermal power and therefore less severe t environmental consequences than a 0.65% Ak/k ejected rod worth at rated ; power. t

Control rod Groups are withdrawn in sequence beginning with Group 1. l Groups 5, 6, and 7 are overlapped 20%. The normal position at power is ,

for Groups 6 and 7 to be partially inserted. l i

             *4ctual operating limits depend on whether or not imcore or excore detectors are 2 sed and their respective instrument and calibration

' errors. The method used to define the operating limits is defined in ; plant operating procedures. l Amendrent No. 52 4Ba i

Table 4.1-1 (Cont.) Channel Descriotion Check Test Calibrate Remarks 47..RCS Subcooling Margin D NA R Monitor

48. Electromatic Relief Valve D NA R Flow Monitor

49. Eluctromatic Relief Block D NA R Valve Position Indicator

                       '50. Pressurizer Safety Valve                                            D                               NA       R Flow Monitor
                       ' 51. Pressurizer Water Level                                            D                               NA       R Indicator

52. Control Room Chlorine Detector D M R .

53. EFW Initiation

a. Manual NA M NA

b. SG Low Level, SGA or B S M R

c. . Low Pressure SGA or B S M R -l

d. Loss of both MFW Pumps S M R and PWR > 107.

                           **-idment No. 75, 7,5%                5   ,# ,71,135                                                                 72b1 I                   .

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                                                                                                                                                                                                                                                  ;  w~<

i Design Basis Accident Leakage Rate. Where (L,) at Pressure P, -l (Lt) Maximum All wable Test u d. age Rate at i Reduced Test Pressure P Under Test l Condition , i (L,9) Maximum allowable operational leakage

rate at pressure P, (L g) Maximum' allowable leakage rate at l pressure P ;

(L,,) Initial Measured Leakage Rate at Pressure l P, .; (L ,) Initial Measured Leakage Rate at Pressure

                                            't                                                      {

(P,) Peak Test Pressure of 59 psig r f (P ) Reduced Test Pressure of 30 psig i 4.4.1 1.3 Conduct of Tests j

a. Leakage rate tests should not be started until essential !

temperature equilibrium has bee-n ettained. Containment !' test conditions should staM11ze for a period of about four hours prior :o the start of a leakage rate test. !

b. The leakage rate test period shall extend'to 24 hours of -;

retained internal prowsure. If it can be demonstrated to the satisfaction of those responsible for the acceptance l of the containment structure that the leakage rate can be accurately determined during a shorter test period, the agreed upon shorter period may be used.

c. Test accuracy shall be verified by supplementary means, such as measuring the quantity of air required to return to the starting point or by imposing a known leak rate to ,

demonstrate the validity of measurements. i

d. Cloaure of reactor building isolation valves for the purpose of the test shall be accomplished by the means provided for normal operation of the valves without preliminary exercise or adjustment. l 4

4.4.1.1.4 Freauenci of Test ! I After the initial preoperational leakage rate test, a set of ! three integrated leak rete teste shall be performed at approximately equal intervals during'each 10 year service- ; period, with the third test of each set coinciding with the [ end of e, sh 10-year service period. The test may coincide i with the plant inservice inspection sLwedown. periods. l l Amendment No. 121 80

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Should the inspection of one of the wires reveal any significant { physical change (pitting or loss of area), additional wires shall ! be removed from the applicable surveillance t-ndons and inspected j to determine the extent and cause of change. The sheathing filler l l will be sampled and inspected for changes in physical appearance. -l (See Applicable Acceptance Criteria in Section 4.4.2.1.3) l l 4.4.2.1.3 Acceptance Criteria ; i The Raactor Building Post Tensioning System shall be considered j acceptable if the following acceptance criteria are met. i

1. Each surveillance tendon has a normalized lif t-off force equaling or exceeding its expected prestress force for the l time of the test. See Figures 4.4.2-1, -2, and -3. If the !

normalized lift-off force of any one tendon in a group lies ; between the expected prestress force and the lower bound [ prestress force, an adjacent tendon on each side shall be ! checked for lift-off force. If both of these tendons are i found acceptable, the surveillance program may proceed i considering the single deficiency as unique and accept.able. l If either of the adjacent tendons is found unacceptable, it j shall be considered evidence of possible abnormal degradatien of j the containment structure. (See TS 6.12.5) [ If the normalized lift-off force of any single tendon lies below the lower bound prestress force, the occurrence should , be considered evidence of possible abnormal degradation of { the containment structure. (See TS 6.12.5) ! i

2. The wires removed from three detensioned surveillance tendons j (one dome, one vertical and one hoop) shall be inspected for i corrosion, cracks, or other damage over the entire length of {

the wire. The presence of abnormal corrosion, cracks, or . other damage shall be considered evidence of possible { abnormal degradation of the containment structure. (See TS ; 6.12.5) { t A minimum of three (3) wire sampics cut from each removed i wire (one from each end and one at mid length) shall be l subjected to a tensile test. Failure of any one of these wire samples to meet a minimum ultimate tensile strength of { 240 kai shall be considered evidence of possible abnormal j degradation of the containment structure. (See TS 6.12.5) j

3. Sheathing Filler material samples from each surveillance I shall be considered acceptable provided the results of the j tests performed on the samples fall within the following ;

limits. j

1. Water Soluble Chlorides less than 10 ppm

2. Water Soluble Nitrates less than 10. ppm j

3. Water Soluble Sulfides less than 10 ppm ;

4. Water Content less than 10% Pry Weight l I

l Amendment No. $$, 118 85a q

I 4.8 EMERGENCY FEEDWATER PUMP TESTING l J Aeolicability Applies to the periodic testing of the turbine and electric motor driven emergency feedwater pumps. l Obiective . ~ To verify that the emergency feedwater pump and associated valves are { operable. Snecification 4.8.1 Each EFW train shall be demonstrated operable: , a) By verifying on a STAGGERED TEST hASIS:

1. at least once per 31 days or upon achievit.c, ot )

shutdown following a plant heatup and prior to ! criticality, that the turbine-driven pump starts, l operates for a minimum of 5 minutes and develops

l a discharge pressure of 11200 psig at a flow of l 1 500 gpm through the test loop flow path. l

2. at least once per 31 days by verifying that the motor driven EFW pump starts, operates for a minimum !

of 5 minutes and develops a discharge pressure of 1 1200 psig at a flow of 1 500 gpm thorough the test j loop flow path. l b) At least once per 31 days by verifying that each valve ! (manual, power operated or automatic) in each EFW flowpath that is not locked, sealed, or otherwise secured ; in position, is in its correct position. ] ~ c) Prior to exceeding 280 F reactor coolant temperature and 'l after any EFW flowpath manual valve alterations by ; verifying that each manual salve in each EFW flowpath which, if mispositioned may degrade EFW operation, is j locked in its correct position. , d) At least once per 92 days by cycling each motor-operated ; valve in each flowpath through at least one complete i l cycle. e) At least once per 18 months by functionally testing each EFW train and:

1) Verifying that each automatic valve in each flowpath actuates automatically to its correct position on receipt of an actuation signal.

Amendment No. 75, 57, PJ, 121 105 i

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     ,   4.30             RADIOLOGICAL ENVIRONMENTAL MONITORING                                            '!

4.30.1 l Radiolorical Environmental Monitorine Procram Descrintion Aonlicability }~ f I Applies at all times. l Obiective l l To provide information on the radiological effects of station operation on the environment. { Specification l 4.30.1.1 The radiological environmental monitoring samples shall be '! collected pursuant to Table 4.30-1 and shall be analyzed f pursuant to the requirements of Tables 4.30-1 and 4.30-2. The j sample locations shall be listed in table 4-1 in the ODCM. l i 4.30.1.2 a. With the radiological environmental monitoring program j not being conducted as specified in Table 4.30-1, prepare ! and submit to the Commission in the Annual Radiological ! Environmental Report a description of the reasons for not conducting the program as. required and the plans for : preventing a recurrence. (Deviations are permitted from ! the required sampling schedule if specimens are not ; obtainable due to hazardous conditions, seasonal ! unavailability, or to malfunction of sampling equipment. { If the latter, every effort shall be made to complete j corrective action prior to the end of the next sampling i period.) {

b. With the level of radioactivity as the result of plant l l effluents in an environmental sampling medium'at one or !

more of the locations specified in Table 4.30-1 exceeding -[ the limits of Table 4.30-3 when averaged over any ; calendar quarter, prepare and submit to the Commission, ! within 30 days from the end of the affected quarter, a ! Special Report which includes an evaluation of any i release conditions, environmental factors or other aspects which caused the limits of Table 4.30-3 to be ! exceeded, and defines the actions taken to reduce } radioactive effluents so that the potential annual dose l to a member of the public is less than the calendar year limits of Specifications 3.25.1.2 and 3.25.2.2. When j more than one of the radionuclides in Table 4.30-3 are [ detected in the sampling medium, this Special Report shall be submitted if: Concentration (1) reporting level (1) , Concentration (2) reporting level (2) + *** 2 1.0 j i When radionuclides other than those in Table 4.30-3 are .! detected and are the result of plant effluents, thi. i . Special Report shall be submitted if the potential annual I dose to a member of the public is equal to or greater '! than the calendar { Amendment No. $$, 118 110ss .j i a b

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year limits of Specifications 3.25.1.2 and 1.25.2.2. This ' I Special Report is not required if the measured level of radioactivity was not the result of plant effluents, -; however, in such an event, the condition shall be reported j and described in the Annual Radiological Environmental e Report. I l

c. With milk or fresh leafy vegetable samples unavailable from !

any of the sample locations required by Table 4.30-1, l Adentify locations for obtaining replacement samples- and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples , were unavailable may then be deleted from the monitoring . program. Identify the causes of the unavailability of ! samples- and identify the new location (s) for . obtaining replacement samples in the next Semiannual Radioactive l Effluent Release Report and also include in the report a revised table for the ODCM reflecting the new location (s).

d. The provisions of Specification 3.0.3 are not applicable.

4.30.1.3 The results of analyses performed on the radfological l environmental monitoring samples shall be summarized in the , Annual Radiological Environmental Report. ! ~ Bases. The radiological monitoring program required by this specification provides ; measurements of radiation and of radioactive materials in those exposure ! pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluents ' monitoring program by verifying that the measurable concentrations of ; radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of_the environmental l exposure pathways. The initially specified monitoring program will be : effective for at least the first three years of commercial operation. [ Following this period, program changes may be initiated based on ; operational experience. ! The detection capabilities required by Table 4.30-2 are state-of-the-art [ for routiae environmental measurements in industrial laboratories. The . LLD's for drinking water meet the requirements of 40 CFR 141. ! P

I

i Amendment No. 55, JJE, 161 110tt L [

1 assumed to be . released into the reactor building through a break in the reactor coolant piping. Subsequent pressure behavior is , determined by the building volume, engineered safety features, and : the combined influence of energy sources and heat sinks. (1) l l 5.2.2 Reactor Building Isolation System Leakage through all fluid penetrations not serving accident-consequence-limiting systems is to be minimized by a double barrier so that no single, credible failure or malfunction of an active component can result in loss-of-isolation or , intolerable leakage. The installed double barriers take the form ' of closed piping systems, both inside and outside the reactor building and various types of isolation valves. (*) l - 5.2.3 Penetration Room Ventilation System This system is designed to collect, control, and minimize the ! release of radioactive material from the reactor building to the i environment in post-accident conditions. It may also operate l intermittently during normal conditions as required to maintain satisfactory temperature in the penetrations rooms. When the . system is in operation, a slightly negative pressure will be {

-{

maintained in the penetration room to assure inleakage. (*) ! , r EEFERENCES: (1) FSAR Section 5.1 (2) FSAR Section 5.2.5 I ,. (3) FSAR Section 6.5 . l i i t l 113 i i i

4 c' . 5.4 NEW AND SPENT FUEL STORAGE FACILITIES { Anolicability . Applies to storage facilities for new and spent fuel assemblies. l Obiective l t i To assure that both new and spent fuel assemblies will be stored in such l a manner that an inadvertent criticality could not occur. l Soccification , 5,4.1 New Fuel Storace

1. Fuel assemblies are stored in racks of parallel rows, having a nominal center to center distance of 21 inches in both ;

directions. This spacing is sufficient to maintain a K of ; lessthan0.9eveniffloodedwithunboratedwater, bas $d'on l i fuel with an enrichment of 3.5 weight percent U235. ; l 2. New fuel may also be stored in the spent fuel pool or in'its j shipping containers. l 5.4.2 Snent Fuel Storace

1. The spent fuel racks are designed and shall be maintained so !

that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) when the ; pool is flooded with unborated wat r. j

2. The spent fuel pool and the new fuel pool racks are designed !'

as seismic. Class I equipment. REFERENCES , FSAR, Section 9.6 , I i a

l t

P 1 ! I i i I i i I j Amendment No. J7, 76 116 1 l l

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. 3 ALTERNATES !

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6.5.1.3 All. alternate members shall be appointed in writing by the PSC ' Chairman to serve on a temporary basis; however, no more than t two alternates shall participate as voting members in PSC [ activities at any one time.

i MEETING FREQUENCY The PSC shall meet at least once per calendar month and as 6.5.1.4 convened by the PSC Chairman or his designated alternate. j OUORUM !

6.5.1.5 The minimum quorum of the PSC necessary for the performance of .' the PSC responsibility and authority provisions of these technical specifications shall consist of the Chairman or his designated alternate and four members including alternates. -! i RESPONSIBILITIES l _ 6.5.1.6 The Plant Safety Committee shall be responsible for: ,

a. Review of 1) all procedures required by Specification j 6.8 and changes in intent thereto, and 2) any other -l proposed procedures or revisions thereto as determined by the General Manager, Plant Operations .

or Plant Manager, ANO-1 to affect nuclear safety. ;

b. Review of all proposed tests and experiments that affect nuclear safety. j

c. Review of all proposed changes to the Appendix "A" ;

Technical Specifications.

d. Review of all proposed changes or modifications to 1 plant systems or equipment that affect nuclear i safety.

e. Investigation of all violations of the Technical Specifications, including the preparation and ,

forwarding of reports covering evaluation and recommendations to prevent recurrence to the Plant Manager, ANO-1, General Manager, Plant Operations and , to the Chairman of the Safety Review Committee. I

f. Review of REPORTABLE EVENTS.

                 .        g. Review of facility operations to detect potential nuclear safety hazards.

Amendment No. $$,99,199,JJf,J15,115, 121 Jf5,147 l I I I i 1

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6.12 REPORTING REOUIREMENTS E 6.12.1 In addition to the applicable reporting requirements of Title ! 10, Code of Federal Regulations, the following identified . > reports shall be submitted to the Administrator of .the ; appropriate NRC Regional Office unless otherwise noted, l I 6.12.2 Routine Reports ! 6.12.2.1 Startup Report ; A summary report of plant startup and power escalation testing shall be l submitted following: 1) receipt of an operating license, 2) amendment to ! the license involving a planned increase in power level, 3) installation ; of fuel that has a different design or has been manufactured by a. different fuel supplier, and 4) modifications that may have significantly 'l 'I altered the nuclear, thermal, or hydraulic performance of the plant. The j report shall address each of the tests identif.ied in the FSAR and shall in ; general include a description of the measured values of the operating ', conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. . Any corrective actions that were required to obtain satisfactory operation ! shall also be described. Any additional specific details required.in ^ license conditions based on other commitments shall be included in this report. t Startup reports _shall be submitted within 1) 90 days following completion - of the startup test program, 2) 90 days following resumption or commencement of commercial power operation, or 3) 9 months'following ! initial criticality, whichever is earliest. If the Startup Report does- ! not cover all three events (i.e.. initial criticality, completion of ! startup test program, and resumption or commencement of commercial power [ operation), supplementary reports shall be submitted at least every three j months until all three events have been completed. ! 6.12.2.2 Occupational Exposure Data Report 1/ i An Occupational Exposure Data Report for the previous calendar year shall j be submitted prior to March 1 of each year. The report shall contain a , tabulation on an annual basis of the number of station, utility and other i personnel (including contractors) receiving exposures greater than 100 ; mrem /yr and their associated man rem exposure according to work and job ' functions, 2/ e.g., reactor operations and surveillance, inservice l 5 inspection, routine maintenance, special maintenance (describe i maintenance), waste processing, and refueling. l

1/ A single submittal may be made for a m,1tiple unit. station. The submittal should combine those sections that are common to all units at the station.

2/ This tabulation supplements the requirements of 20.407 of 10 CFR Part

20. ]

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i Amendment No. 7, 17, 82 140 : I I

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ML20044B693

Text

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