ML18344A285
| ML18344A285 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 01/20/1998 |
| From: | Consumers Energy Co, (Formerly Consumers Power Co) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18344A285 (218) | |
Text
{{#Wiki_filter:IMPROVED TECHNICAL SPECIFICATIONS ALI SADES UCLEAR , ,, LANT l volume 1a SECTION 3.9 Consumers Energy
PALISADES NUCLEAR*PLANT CONSUMERS ENERGY Docket 50-255 Conversion to Improved Technical Specifications License DPR-20 INTRODUCTION: SECTION 3.9. REFUELING OPERATIONS A. ARRANGEMENT AND CONTENT OF THIS SECTION OF THE CHANGE REQUEST This Section of the Technical Specification Change Request (TSCR) proposes changes to those Palisades Technical Specifications addressing REFUELING OPERATIONS, and the associated surveillance requirements in Section 4. These changes are intended to result in requirements which are appropriate for the Palisades Nuclear Plant, but closely emulate those of the Standard Technical Specifi~ations, Combustion Engineering Plants, NUREG 1432, Revision 1, Section 3.9. This discussion and its supporting information frequently refer to three sets of Technical Specifications, and to two groups of discussions associated with the proposed changes; the following abbreviations are used for clarity and brevity: CTS - The Palisades Current Technical Specifications, ITS - The Palisades Improved Technical Specifications, ISTS - NUREG 1432, Revision 1. DOC - Discussions of Change; these discussions explain and justify the differences between the requirements of CTS and ITS. *
- JFD 1.
- Justifications for Deviation; these discussions explain the differences between the requirements of the ITS and the ISTS.
Six attachments are provided to assist the reviewer: Proposed ITS Section 3.9 pages
- 2. Proposed ITS Section 3.9 Bases
- 3. A set of all those CTS pages which contain requirements associated with those in ITS Section 3.9, marked up to show the changes from CTS to ITS, and arranged by specification in the order in which the requirements occur in ITS. This attachment also includes a DOC for each change.
Each change from CTS to ITS is classified in the following categories: ADMINISTRATIVE - A change which is editorial in nature, which only involves movement of requirements within the TS without affecting their technical content, or clarifies CTS requirements. MORE RESTRICTIVE - A change which only adds new requirements, or which revised an existing requirement resulting in additional operational restrictions. RELOCATED - A change which only moves requirements, not meeting the 10 CFR 50.36(c)(2)(ii) criteria, from the CTS to the Operating Requirements Manual (which has been included in the FSAR by reference) . 1
INTRODUCTION: SECTION 3.9. REFUELING OPERATIONS
- A. ARRANGEMENT AND CONTENT OF THIS SECTION OF THE CHANGE REQUEST (continued)
LESS RESTRICTIVE - REMOVAL OF DETAIL (LA) - A change in which certain details from otherwise retained Specifications are removed from the ITS and placed in the Bases, FSAR, or other licensee controlled documents. LESS ~~STRICTIVE - A change which deletes any existing requirement, or which revises any existing requirement resulting in reduced operational restrictions.
- 4. No Significant Hazards Analyses for the changes from CTS to ITS.
An individual No Significant Hazards Analysis is provided for each Less Restrictive change; generic No Si gni fi cant Hazards Ana 1yses are p-rovi ded for eaGh of the other categories of change.
- 5. ISTS Section 3.9, Specifications and Bases, marked to show the differences between ISTS and ITS.
- 6. JFDs for the differences between ISTS and ITS.
B. REFERENCE DOCUMENTS
- This section of the TSCR is based on the following source documents:
1. 2. CTS as revised through Amendment 178. The following TSCRs which are currently under review by the NRC:
- a. Electrical, initially submitted on December 27, 1995.
- 3. ISTS, as revised by Industry Generic Changes (TSTF) approved as of October 15, 1997.
- 4. The following changes to ISTS which are currently under review by the NRC:
- a. None *
- 2.
INTRODUCTION: SECTION 3. 9. REFUELING OPERATIONS * '
- C. THE UNIQUE PALISADES NUCLEAR PLANT FEATURES AFFECTING THIS SECTION Palisades has several physical, analytical, and administrative features which differ from those newer CE plants upon which the ISTS were based. Palisades was the first CE plant designed and built. Its design and licensing preceded the issuance of the General Design Criteria so that, in some aspects, its physical systems ar~__ not like those of newer plants; its Technical Specifications preceded the issuance of Standard Technical Specifications (STS) so that LCOs, Actions, and Surveillance Requirements are not coordinated as they would be for a STS plant.
Palisades has purchased all its core reloads from Siemens Power Corporation (or its predecessors), therefore, reload analyses and the associated core physics parameters, as well as certain Safety Analyses are not like those plants using all CE fuel and analyses as were modeled in the ISTS. The following unique features affect the proposed requirements of this section:
- 1. The Containment Equipment Hatch, at Palisades, opens directly into the fuel storage area. This feature provides an area with a controlled ventilation path outside the Equipment Hatch. The ITS retain the CTS allowance to perform Core Alterations with the Equipment Hatch open.
- 2. The Shutdown Cooling System and its support systems do not all have two fully independent trains. Several exceptions which allow temporary interruption of shutdown cooling, or of mixing flow through the core have been retained to allow performance of testing or maintenance.
D. THE DIFFERENCES BETWEEN CTS "OPERATING CONDITIONS" AND ITS "MODES" The CTS definitions of plant operating conditions have been replaced with the operation Mode definitions used in ISTS. In several instances the name for aCTS defined 11 operating condition 11 is the same as that for an ISTS 11 Mode, 11 but the definition differs. CTS contain the following definitions for operating conditions:
- 1. The POWER OPERATION condition shall be when the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of RATED POWER.
- 2. The HOT STANDBY condition ~hall be when T~e is greater than 525°F and any of the CONTROL RODS are withdrawn and the neutron flux power range instrumentation indicates less than 2%.of RATED POWER. *
- 3. The HOT SHUTDOWN condition shall be when the reactor is subcritical by an amount greater than or equa 1 to the margin as specified in Tecl:mi ca 1 Speci fi ca ti on 3 .10 and Tave is greater than 525°F.
- 4. The COLD SHUTDOWN condition shall be when the primary coolant is at SHUTDOWN
- BORON CONCENTRATION and Tave is 1ess than 210°F .
3
INTRODUCTION: SECTION 3.9. REFUELING OPERATIONS
- D.. THE DIFFERENCES BETWEEN CTS "OPERATING CONDITIONS" AND ITS
- 5. The REFUELING SHUTDOWN condition shall be when the primary coolant is at REFUELING BORON CONCENTRATION and Tave is 1ess than 210°F.
"MODES,~ (continued)
ITS contain the following definition table for Modes: MODE TITLE REACTIVITY CONDITION
% RATED THERMAL - AVERAGE PRIMARY (k.,,) POWER(a) COOLANT TEMPERATURE (°F) 1 Power Operation ~ 0.99 > 5 NA 2 Startup ~ 0.99 ,:; 5 NA 3 Hot Standby < 0.99 NA ~ 300 4 Hot Shutdown lbl < 0.99 NA 300 >. Tave > 200 5 Cold Shutdown<bl < 0.99 NA ,:; 200 Refuel ing" 1 6 NA NA NA (a) Excluding decay heat.
(b) All reactor vessel head. closure bolts fullY:tensicined. (c) One or more reactor vessel head closure bolts less than fully tensioned. E. MODE CHANGES USING CTS "OPERATING CONDITIONS" VERSUS ITS "MODES"
- 1. CTS "Power Operation" is essentially equivalent to ITS "MODE 1." Each represents a condition with the reactor critical and the turbine generator in operation. The only effective difference is the power level which separates that Condition or Mode from the next lower one. During plant startup, the plant must meet all CTS "Power Operation" or ITS "MODE 1" LCOs before the turbine generator is placed on the line; similar, during plant shutdown, the_
plant exits CTS "Power Operation" or ITS "MODE 1" when the turbine generator is no longer in service. Therefore, this change in definition will have no operational effect.
- 2. CTS "Hot Standby" is similar to ITS "MODE 2." Each represents a condition with the reactor critical, or nearly so, and the turbine generator shut down.
During plant startup, the plant must meet all CTS "Hot Standby" cir ITS "MODE 2" LCOs before a reactor startup is started; during plant shutdown, the plant exits CTS "Hot Standby" or ITS "MODE 2" when the reactor is shutdown. CTS action statements requiring that the plant be placed in 11 Hot Standby" are effectively equivalent to ITS Actions requiring the plant be placed in "MODE 2. 11 Therefore, this change in definition wi 11 have no opera ti ona 1 effect. 4
INTRODUCTION: SECTION 3. 9, REFUELING OPERATIONS*"*-,, ;
- E. MODE CHANGES USING CTS "OPERATING CONDITIONS" VERSUS ITS "MODES" ~
3. (continued) CTS "Hot Shutdown" and ITS "MODE 3" are similar at their upper te~perature boundary. During plant shutdown, the plant exits CTS "Hot Standby" or ITS "MODE 2" when the reactor is shutdown. CTS action statements requiring that the plant be placed in "Hot Shutdown" are effectively equivalent to ITS Action_s requiring the plant be placed in "MODE 3." CTS "Hot Shutdown" and ITS "MODE 3 are quite different at their lower temperature boundary; CTS 11 "Hot Shutdown" is exited when Tave drops below 525°F, ITS "MODE 3" is not exited until Tave drops below 300°F.
- 4. CTS does not provide a defined term for the condition when Tave is between 525°F and 210°F (the upper bound for CTS "Cold Shutdown").
- 5. CTS "Cold Shutdown" is essentially equivalent to ITS "MODE 5." Each represents a condition with Tave below boiling. There is no technical significance to the difference between the CTS 210°F and the ITS 200°F.
CTS action statements requiring that the plant be placed in "Cold Shutdown" are effectively equivalent to ITS Actions requiring the plant be placed in "MODE 5." Therefore, this change in definition will have no operational effect. *
- 6. CTS "Refueling Shutdown" is essentially equivalent to ITS "MODE 6. Each, 11 when taken with other definitions and LCO requirements, represents a
- F.
condition with the reactor at least 5% shutdown. Therefore, this change in definition will have no operational effect. THE MAJOR CHANGES FROM CTS (as modified by pending TSCRs) TO ITS
- 1. The ITS require Refueling Boron Concentration to be maintained throughout MODE 6 operation, rather than the*CTS~requirement which is applicable during actual core alterations.
- 2. The ITS contain an additional Specification on Reactor Cavity water level.
This is a new Specification for Palisades. The requirement for reactor cavity water level currently exists only in plant procedures. G. THE MAJOR DIFFERENCES BETWEEN ITS AND ISTS
- 1. The CTS definition of Refueling Boron Concentration has been retained; there is no equivalent definition in the Palisades COLR nor in ISTS. *ITS 3.9.1 requires maintaining the PCS and Refueling Cavity at Refueling Boron Concentration (as defined in Chapter 1.0) rather than maintaining it at a boron concentration listed in the COLR. The term "Refueling Boron Concentration" is also used in ITS Section 3.3. *
- 5
INTRODUCTION: SECTION 3.9, REFUELING OPERATIONS .
- G. THE MAJOR DIFFERENCES BETWEEN ITS AND ISTS 2.
(continued) The CTS allowance to conduct Core Alterations with the containment equipment hatch open has been retained; the existing requirements allow both doors of the personnel hatch and the equipment hatch to be open during Core Alterations provided the spent fuel pool ventilation system and charcoal filter--.is in operation. The equipment hatch opens directly into the fuel storage building.
- 3. Several CTS exceptions allowing temporary interruption of core cooling or of mixing flow through the reactor core have been retained. Si~il~r exceptions appear in the PCS Specifications whi~h deal with PCS circulation during shutdown conditions *
- 6
ATTACHMENT 1 PALISADES NUCLEAR PLANT
- SECTION 3.9 - REFUELING OPERATIONS PROPOSED TECHNICAL SPECIFICATIONS
Boron Concentration 3.9.1
- . 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Primary Coolant System and the refueling cavity shall be maintained at the REFUELING BORON CONCENTRATION.
APPLICABILITY: MODE 6. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration A.1 Suspend CORE Immediately not within limit. AL TERATI ONS. AND A.2 Suspend positive Immediately reactivity additions. AND A.3 Initiate action to Immediately restore boron concentration to with i n 1i mi t.
. I I
I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 9 .1.1 Verify boron concentration is at the 72 hours REFUELING BORON CONCENTRATION .
- Palisades Nuclear Plant 3.9.1-1 Amendment No. 01/20/98
Nuclear Instrumentation 3.9.2
- 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation LCO 3.9.2 Two source range channels shall be OPERABLE.
APPLICABILITY: MODE 6. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One source range A.1 Suspend CORE Immediately channel inoperable. ALTERATIONS. AND A.2 Suspend positive Immediately reactivity additions.
- B. Two source range channels inoperable.
B.1 AND Initiate action to restore one source range channel to OPERABLE status. Immediately. B.2 Perform SR 3.9.1.1 Once.per (PCS boron 12 hours concentration verification) .
- Palisades Nuclear Plant 3.9.2-1 Amendment No. 01/20/98
Nuclear Instrumentation 3.9.2
- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK. 12 hours SR 3.9.2.2 Perform CHANNEL CALIBRATION. 18 months
- Palisades Nuclear Plant 3.9.2-2 Amendment No. 01/20/98
Containment Penetrations 3.9.3
- 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:
- a. The equipment hatch closed and held in place by four bolts;
----------------------------NOTE----------------------------
The equipment hatch is only required to be closed when the Fuel Handling Area Ventilation System is not in compliance with LCO 3.7.12, 11 Fuel Handling Area Ventilation System. 11
- b. One door in the personnel air lock closed;*
----------------------------NOTE----------------------------
One door in the personnel air lock is only required to be closed when the equipment hatch is closed.
- c.
d. One door in the emergency air lock closed; and Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
- 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or
- 2. capable of being closed by an OPERABLE Refueling Containment High Radiation Initiation signal.
APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment .
- Palisades Nuclear Plant 3.9.3-1 Amendment No. 01/20/98
Containment Penetrations 3.9.3
- ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend CORE Immediately containment ALTERATIONS.
penetrations not in required status. AND A.2 Suspend movement of Immediately irradiated fue 1 assemblies within containment. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
- SR 3.9.3~1 Verify each required containment penetration is in the required status.
7 days SR 3.9.3.2 -------------------NOTE-------------------- Only required for unisolated containment penetrations. Verify each required automatic isolation 18 months valve closes on an actual or simulated Refueling Containment High Radiation signal .
- Palisades Nuclear Plant 3.9.3-2 Amendment No. 01/20/98
- SOC and Coolant Circulation - High Water Level 3.9.4
- .3.9 REFUELING OPERATIONS 3.9.4 Shutdown Cooling (SOC) and Coolant Circulation - High Water Level LCO 3.9.4 One SOC train shall be OPERABLE and in operation.
----------------------------NOTES---------------------------
- 1. The required SOC train may not be in operation for
~ 1 hour per 8 hour period, provided no operations*
are permitted that would cause reduction of the , Primary Coolant System boron concentration. *
- 2. The required SOC train may be made inoperable for
~ 2 hours per 8 hour period for testing or maintenance, provided one SOC train is in operation .
providing flow through the reactor core, and core outlet temperature is ~ 200°F. APPLICABILITY: MODE 6 with the refueling cavity water level ~ 647 ft elevation.
* =AC=T=I=ON=S=================;r======================::::::;:::============
CONDITION REQUIRED ACTION COMPLETION TIME A. One required SOC train A.1 Initiate action to Immediately_ inoperable or not in
- restore SOC train to operation. OPERABLE status and operation.
A.2 Suspend operations Immediately involving a reduction in primary coolant boron concentration. (continued)
- Palisades Nuclear Plant 3.9.4-1 Amendment No. 01/20/98
SOC and Coolant Circulation - High Water Level 3.9.4
- _AC_T_IO_N_S______________--r----------------------"T----------~
CONDITION REQUIRED ACTION . COMPLETION TIME A. (continued) A.3 Suspend loading Immediately irradiated fuel assemblies in the core. A.4 Close all containment 4 hours penetrations providing direct access from containment atmosphere to outside atmosphere.
- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify one SOC train is in operation and 12 hours circulating primary coolant at a flow rate of ~ 1000 gpm .
- Palisades Nuclear Plant 3.9.4-2 Amendment No. 01/20/98
SOC and Coolant Circulation - Low Water Level 3.9.5
- 3.9 REFUELING OPERATIONS
-3.9.5 Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level LCO 3.9.5 Two SOC trains shall be OPERABLE, and one SOC train shall be i.n opera ti on.
APPLICABILITY: MODE 6 with the refueling cavity watef level < 647 ft elevation. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One SOC train A.1 Initiate action to Immediately inoperable. restore SOC train to OPERABLE status. OR
- A.2 Initiate action to establish the refueling cavity water level ~ 647 ft elevation .
Immediately
- Palisades Nuclear Plant 3.9.5-1 Amendment No. 01/20/98
SOC and Coolant Circulation - Low Water Level 3.9.5 CONDITION REQUIRED ACTION . COMPLETION TIME B. No SOC train OPERABLE B.1 Suspend operations Immediately or in operation. involving a reduction in primary coolant boron concentration. AND B.2 Initiate action to Immediately restore one SOC train to OPERABLE status and to operation. AND B.3 Initiate action to Immediately close all containment penetrations
- providing direct access from containment atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS
- SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one SDC train is in operation*and 12 hours circulating primary coolant at a flow rate of ~ 1000 gpm.
SR 3.9.5.2 Verify correct breaker alignment and 7 days indicated power available to the required SDC pump that is not in operation .
- Palisades Nuclear Plant 3.9.5-2 Amendment No. 01/20/98
Refueling Cavity Water Level 3.9.6
- 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCD 3.9.6 The refueling cavity water level shall be maintained
~ 647 ft elevation.
APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water A. l Suspend CORE Immediately level not within ALTERATIONS. 1imi t. AND A.2 Suspend movement of Immediately irradiated fue 1 assemblies within containment. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is 24 hours
~ 647 ft elevation .
- Palisades Nuclear Plant 3.9.6-1 Amendment No. 01/20/98
ATTAC1'ENT 2 PALISADES NUCLEAR PLANT
- SECTION 3.9 - REFUELING OPERATIONS PROPOSED BASES
Boron Concentration B 3.9.1
- B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration BASES BACKGROUND The limit on the boron concentrations of the Primary Coolant System (PCS), and refueling cavity during refueling ensures -
that the reactor remains subcritical during MODE 6. The refueling operations boron concentration is the soluble boron concentration in the coolant in each of these volumes having direct access to the reactor core d~ring refueling. The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The REFUELING BORON CONCENTRATION limit is defined in Section 1.1, "Definitions." Plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of keff ~ 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity) allowed by plant procedures. During evolutions where plant procedures allow manipulation of control rods or where conditions could
- result in inadvertent control rod withdrawal, such as reactor vessel head removal, the boron concentration must be sufficient to assure that ~ff will remain ~ 0.95 without taking credit for the negative reactivity provided by the control rods (i.e., assuming all rods fully withdrawn).
During evolutions where the control rods are inserted, plant procedures do not allow manipulation of control rods, and conditions do not exist that could result in inadvertent rod withdrawal, such as MODE 6 operations with the Upper Guide Structure in place (other than during head removal). Therefore, credit may be taken for the negative reactivity provided by the control rods when determining the boron concentration necessary to assure that keff wi 11 remain
~ 0.95.
The Palisades Nuclear Plant design criteria requires that
*two independent reactivity control systems of different design principles be provided (Ref. 1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. The Chemical and Volume Control System (CVCS) System is capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration .
- Palisades Nuclear Plant B 3.9.1-1 01/20/98
Boron Concentration B 3.9.1
- BASES BACKGROUND (continued)
The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for refueling. After the PCS is cooled and depressurized the vessel head is unbolted and the head is removed. The refueling cavity is then flooded with borated water from the safety injection refueling water tank into the open reactor vessel by gravity feeding or by the use of the spent fuel cooling, safety injection pumps, or charging pumps. The pumping action of the SOC System in the PCS and the natural circulation due to thermal driving head in the reactor vessel mix the added concentrated boric acid with the water in the refueling cavity. The SOC System is in operation during refueling (see LCO 3.9.4, "Shutdown Cooling (SOC) and Coolant Circulation - High Water Level," and LCO 3.9.5, "Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level") to provide forced circulation in the PCS and to assist in maintaining the REFUELING BORON CONCENTRATION in the PCS, and the refueling cavity .
- APPLICABLE SAFETY ANALYSES During refueling operations, the reactivity condition of the core is consistent with the initial conditions assumed for the boron dilution accident analysis and is conservative for MODE 6. The REFUELING BORON CONCENTRATION limit is based on the core reactivity and includes an uncertainty allowance.
The required boron concentration and the plant refueling procedures that demonstrate the correct fuel loading plan (including full core mapping) ensure the krl, of the core will remain ~ 0.95 during the refueling operation. Hence, at least a 5% ~k/k margin of safety is established during refueling. During refueling, the water volume in the spent fuel pool, the transfer canal, the refueling cavity, and the reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these volumes. The limiting boron dilution accident analyzed occurs in MODE 5 (Ref. 2). A detailed discussion of this event is provided in B 3.1.1, "SHUTDOWN MARGIN." Boron concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2) .
- Palisades Nuclear Plant B 3.9.1-2 . 01/20/98
Boron Concentration B 3.9.1
- BASES LCO The LCO requires that a minimum boron concentration be maintained in the PCS, and refueling cavity while in MODE 6.
The boron concentration limit specified ensures a core keff of~ 0.95 is maintained during fuel handling operations. Violation of the LCO could lead to an inadvertent criticality. APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a k~, ~ 0.95. Above MODE 6, LCO 3.1.1, "SHUTDOWN MARGIN (SDM}," ensure that an adequate amount of negative reactivity is available to shut down the reactor and to maintain it subcritical. ACTIONS A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the plant in compliance with the LCO. If the boron concentration of any coolant volume in the PCS or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately. Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position, or normal cooldown of the coolant volume for the purpose of system temperature control. A.3 In addition to immediately suspending CORE ALTERATIONS or positive reactivity additions, boration to restore the concentration must be initiated immediately. In determining the required combination of boration flow rate and concentration, there is no unique design basis event that must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for plant conditions .
- Palisades Nuclear Plant B 3.9.1-3 01/20/98
Boron Concentration B 3.9.1
- *BASES ACTIONS A.3 (continued)
Once boration is initiated, it must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration. SURVEILLANCE SR 3. 9 .1.1 REQUIREMENTS This SR ensures the coolant boron concentration in the PCS and the refueling cavity is within the limit. The boron concentration of the coolant in each volume is determined periodically by chemical analysis. A minimum Frequency of once every 72 hours is therefore a reasonable amount of time to verify the boron concentration of representative samples. The Frequency is based on operating experience, which has shown 72 hours to be adequate .
- REFERENCES 1.
2. FSAR, Section 5.1 FSAR, Section 14.3
- Palisades Nuclear Plant B 3.9.1-4 01/20/98
Nuclear Instrumentation B 3.9.2
- B 3.9 REFUELING OPERATIONS B 3.9.2 Nuclear Instrumentation BASES BACKGROUND The source range channels (NI-01/03 and NI-02/04) are used during refueling operations to monitor the core reactivity condition. The installed source range channels are part of the Nuclear Instrumentation System. These detectors are located external to the reactor vessel and detect neutrons leaking from the core. The use of portable detectors is permitted, provided the ,LCO requirements are met.
The installed source range channels utilize fission detectors operating in the proportional region of the gas filled detector characteristic curve. The detectors monitor the neutron flux in counts per second. The instrument range covers five decades of neutron flux (1E+5 cps). The detectors provide continuous visual and audible indication in the control room to alert operators to a possible dilution accident. The Nuclear Instrumentation System is
- designed in accordance with the criteria presented in Reference 1.
If used, portable detectors should be functionally equivalent to the installed source range channels. APPLICABLE Two OPERABLE source range channels are required to provide SAFETY ANALYSES a signal to alert the operator to unexpected changes in core reactivity such as by a boron dilution accident or an improperly loaded fuel assembly. The safety analysis of the uncontrolled boron dilution accident is described in Reference 2. The analysis of the uncontrolled boron dilution accident shows that normally available SHUTDOWN MARGIN would be reduced, but there is sufficient time for the operator to take corrective actions. Nuclear Instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2) .
- Palisades Nuclear Plant B 3.9.2-1 01/20/98
Nuclear Instrumentation B 3.9.2
- BASES LCO This LCO requires two source range channels OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE, each channel must provide visual indication and at least one of the two channels must provide an audible count rate function in the control room. Therefore, with no audible count rate function from at least one channel, both source range channels would be inoperable.
APPLICABILITY In MODE 6, the source range channels must be OPERABLE to detect changes in core reactivity.
- There is no other direct means available to check core reactivity levels.
In MODES 2 3, 4, and 5 with no more than one control rod 9 capable of withdrawal, the installed source range channels are required to be OPERABLE by LCO 3.3.9, "Neutron Flux Monitoring Channels." In MODE l, one source range channel
*.is required by LCO 3.~.8, "Alternate Shutdown System."
- ACTIONS A.1 and A.2 With only one source range channel OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and positive reactivity addition~ must be suspended.immediately. Performance of Required Action A.l shall not preclude completion of movement of a component to a safe position or normal cooldown of the coolant volume for the purpose of system temperature control; With no source range channel OPERABLE, action to restore a*
channel to OPERABLE status shall be initiated immediately. Once initiated, action shall be continued until one source range channel is restored to OPERABLE status. Palisades Nuclear Plant B 3.9.2-2 01/20/98
Nuclear Instrumentation B 3.9.2
- BASES ACTIONS (continued)
B.2 With no source range channel OPERABLE, there is no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the source range channel are OPERABLE. This stabilized condition is determined by performing SR 3.9.1.1 to verify that the required boron concentration exists. The Completion Time of once per 12 hours is sufficient *to obtain and analyze a primary coolant and refueling cavity sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour Frequency is reasonable, considering the low . probability of a change in core reactivity during this per*i od. SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions, but does not require the two source range channels to have the same reading. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.
- The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.9 .
- Palisades Nuclear Plant B 3.9.2-3 01/20/98
Nuclear Instrumentation B 3.9.2
- BASES SURVEILLANCE REQUIREMENTS SR 3.9.2.2 (continued) SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION every 18 months. The CHANNEL CALIBRATION for the source range neutron flux monitors consists of obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer*s data.
The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components usually pass the Surveillance when performed on the 18 month Frequency. REFERENCES 1. FSAR, Section 7 .6
- 2. FSAR, Section 14.3
- Palisades Nuclear Plant B 3.9.2-4 01/20/98
Containment Penetrations B 3.9.3
- B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGROUND During CORE ALTERATIONS or movement of fuel assemblies within containment with irradiated fuel in containment, a release of fission product radioactivity within the containment will be restricted from escaping to the environment when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3. 6.1, 11 Contai nment. II In MODE 6, the potential for containment pressurization as a result.
of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as 11 containment closure 11 rather than 11 containment OPERABILITY. 11 Containment closure means that all potential escape paths are filtered, closed or capable of being closed. Since there is no potential for containment pressurization, the 10 CFR 50, Appendix J leakage criteria and tests are not required. In MODE 5, no accidents are assumed which will result in a release of radioactive material to the containment atmosphere. Therefore, no requirements are stipulated for containment penetrations in MODE 5. The containment serves to contain fission product . radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. Additionally, the containment structure provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions. The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out o~ containment. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment with the equipment hatch closed, the hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced .
- Palisades Nuclear Plant B 3.9.3-1 01/20/98
Containment Penetrations B 3.9.3 .* _BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~--'-~ BACKGROUND During CORE ALTERATIONS or-movement of irradiated fuel (continued) assemblies within containment with the equipment hatch removed, the OPERABILITY requirements of the Fuel Handling Area Ventilation System must be met. These OPERABILITY requirements are provided in LCO 3.7.12, "Fuel Handling Area Ventilation System." The containment air locks, which are also part of the ,. containment pressure boundary, provide a means for personnel access during MODE~ 1, 2, 3, and 4 operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has* a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPE~ABILITY is required. During periods of shutdown. when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain closed. An exception, hciwever, is provided for the personnel air lock. It is acceptable to have both doors of the personnel air lock opened simultaneously provided the equipment hatch is opened. The requirements on containment penetration closure ensure that a release of fission product radioactivity within* containment will be restricted from escaping to the environment.* The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel handling accident during refueling. The Containment Purge and Vent System includes a 12 inch purge penetration and two 8 inch exhaust penetrations. During MODES 1, 2, 3, and 4, the valves in the purge and vent penetrations are secured in the closed position and venting the containment is accomplished using the Clean Waste Receiving Tank (CWRT) vent line~ The two valves in the CWRT vent line penetration are closed automatically by a Containment High Radiation signal. Neither the Containment Purge and Vent System, nor the CWRT vent line is subject to a Specification in MODE 5.
- Palisades Nuclear Plant* B 3.9.3-2 01/20/98
Containment ' Penetrations B 3.9.3
- _BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~
BACKGROUND In MODE 6, large air exchanges are necessary to conduct (continued) refueling operations. The Purge and Vent System is used for this purpose. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment with either the Containment Purge and Vent System in operation, or the CWRT aligned for containment venting, the associated isolation valves must be capable of being closed by an OPERABLE channel of radiation instrumentation required by LCO 3.3.6, "Refueling Containment High Radiation Instrumentation." Other containment penetrations that provid~ direct access from containment atmosphere to outside atmosphere that are not capable of being closed by an OPERABLE Refueling Containment High Radiation signal must be isolated on at least one side. Containment penetrations "that provide direct access from containment atmosphere to outside atmosphere" are those which would allow passage of air containing radioactive particulates to migrate from inside the containment to the atmosRhere outside the containment even though no measurable differential pressure existed. Specifically, they do not include penetrations which are filtered, or penetrations whose piping is filled with liquid. Isolation may be achieved by a manual or automatic isolation valve, blind flange, or equivalent. Equivalent isolation methods, authorized under the provisions of 10 CFR 50.59, may include use of a material that can provide a temporary, atmospheric pressure ventilation barrier for the other containment penetrations during fuel movements. APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel SAFETY ANALYSES assemblies within containment, the most severe radiological consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 1). The requirements of LCO 3.9.6, "Refueling Cavity Water Level," (and the minimum decay time of 48 hours required by the Operating Requirements Manual) prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are less than the guideline values specified in 10 CFR 100 .
- Palisades Nuclear Plant B 3.9.3-3 01/20/98
Containment Penetrations B 3.9.3
- _BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~~~~~-
APPLICABLE Containment penetration isolation is not required by the SAFETY ANALYSES fuel handling accident to maintain offsite doses within the guidelines of 10 CFR 100, but operating experience indicates that containment isolation provides significant reduction of the resulting offsite doses. Therefore, the Containment Penetrations satisfy the requirements of Criterion 4 of 10 CFR 50.36(c)(2). LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment .. The LCO requires the equipment hatch, air locks and any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment penetrations. For the OPERABLE containment penetrations, this LCO ensures that these penetrations are isolable by the Refueling Containment High Radiation instrumentation. The OPERABILITY requirements for this LCO do not assume a specific closure time for the valves in these penetrations since the accident
- analysis makes no specific assumptions about containment closure time after a fuel handling accident.
LCO 3.9.3.a is modified by a Note which allows the equipment hatch to be opened if the Fuel Handling Area Ventilation System is in compliance with LCO 3.7.12. LCO 3.9.3.b is modified by a Note which allows both doors of the personnel air lock to be simultaneously opened provided the equipment hatch is opened. With both doors in the personnel air lock opened and the equipment hatch opened, the Fuel Handling Area Ventilation System maintains the atmosphere in the spent fuel pool area at a negative pressure relative to the auxiliary building (adjacent to the personnel air lock) and containment building. In the event of a fuel handling accident inside containment; any radioactivity released to the containment atmosphere will either remain in the containment or be filtered through the Fuel Handling Area Ventilation System. As such, with the equipment hatch removed, and both personnel air lock doors opened, the consequences of a fuel handling accident in containment would not exceed those calculated for a fuel handling accident in the sperit fuel pool area *
- Palisades Nuclear Plant B 3.9.3-4 01/20/98
Containment Penetrations B 3.9.3
- _BA_S_E_S~~~~~~~~~~~~~~~~~~~~~~---,,-~~~
APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1, "Containment." In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status. ACTIONS A.1 and A.2 With the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere not in the required status, including the Refueling Containment High Radiation instrumentation not capable of automatic actuation when the purge and exhaust valves are open, the plant must be placed in a condition in which containment closure is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position. SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the valves in unisolated penetrations which provide a direct path from the containment atmosphere to the outside atmosphere will demonstrate that the valves are not blocked from closing. Also, the Surveillance will demonstrate that each valve operator has motive power, which will ensure each valve is capable of being closed by an OPERABLE Refueling Containment High Radiation signal .
- Palisades Nuclear Plant B 3.9.3-5 01/20/98
Containment. Penetrations B 3.9.3
- BASES
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-
SUR VEIL LANCE SR 3.9.3.1 (continued) REQUIREMENTS The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. As such, this Surveillance helps ensure that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in an excessive release of fission product radioactivity to the environment. SR 3.9.3.2 This Surveillance demonstrates that each automatic isolation valve providing direct access from the containment atmosphere to the outside atmosphere valve actuates to its isolation position on an actual or simulated high radiation signal. The SR is modified by a Note which requires only the valves in unisolated penetrations to be tested. The 18 month
- Frequency maintains consistency with other similar ESFAS instrumentation and valve testing requirements. LCO 3.3.6, "Refueling Containment High Radiation Instrumentation,"
requires a CHANNEL CHECK every 7 days9 a CHANNEL FUNCTIONAL TEST every 31 days and a CHANNEL CALIBRATION every 18 months to ensure the channel OPERABILITY during refueling operations. These surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment. REFERENCES 1. FSAR, Section 14.19
- Palisades Nuclear Plant B 3.9.3-6 01/20/98
SOC and Coolant Circulation - High Water Level
. B 3.9.4
- B 3.9 REFUELING OPERATIONS B 3.9.4 Shutdown Cooling (SOC) and Coolant Circulation - High Water Level BASES BACKGROUND The purposes of the SOC System in MODE 6 are to remove decay heat and sensible heat from the Primary Coolant System (PCS), as required by the Palisade Nuclear Plant design, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification (Ref. 1). Heat is removed from the PCS by circulating primary coolant through the SOC heat exchanger(s), where the heat is transferred to the Component Cooling Water System via the SOC heat exchanger(s). The coolant is then returned to the PCS via the PCS cold leg(s). Operation of the SOC System for normal cooldown or decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of *primary coolant through the SOC heat exchanger(s) and bypassing the heat exchanger(s). Mixing of the primary coolant is maintained by this continuous circulation of primary coolant through the SOC System.
APPLICABLE If the primary coolant temperature is not maintained below SAFETY ANALYSES 200°F, boiling of the primary coolant could result. This could lead to inadequate cooling of the reattor fuel due to the resulting loss of coolant in the reactor vessel. Additionally, boiling of the primary coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of the boiling activity, and because of the possible addition of water to the reactor vessel with a lower boron concentration than is required to keep the reactor subcritical. The loss of primary coolant and the reduction of boron concentration in the primary coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. One train of the SOC System is required to be in operation in MODE 6, with the refueling cavity water level greater than or equal to the 647 ft elevation, to prevent this challenge. The LCO allows the removal of an SOC train from operation for short durations under the condition that the boron concentration is not diluted .
- Palisades Nuclear Plant B 3.9.4-1 01/20/98
SDC and Coolant Circulation - High Water Level B 3.9.4
- BASES APPLICABLE SAFETY ANALYSES This conditional allowance does not result in a challenge to the fission product barrier.
(continued) SOC and Coolant Circulation-High Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2). LCO Only one SOC train is required for decay heat removal in MODE 6, with the refueling cavity water level greater than or equal to the 647 ft elevation. Only one SOC train is required because the volume of water above the reactor . vessel flange provides backup decaj heat removal capability. At least one SDC train must be OPERABLE and in operation to provide:
- a. Removal of decay heat;
- b. Mixing of borated coolant to minimize the possibility of a criticality; and
- c. Indication of reactor coolant temperature .
An OPERABLE SOC train consists of an SOC pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the PCS temperature. The flow path starts in one of the PCS hot legs and is returned to at least one PCS cold leg. The LCO is modified by two Notes. Note 1 allows the required operating SOC train to not be in operation for up to 1 hour in each 8 hour period, provided no operations are permitted that would cause a reduction of the PCS boron concentration. Boron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core mapping or alterations in the vicinity of the reactor vessel hot leg nozzles, and PCS to SOC isolation valve testing *
- Palisades Nuclear Plant B 3.9.4-2 01/20/98
SOC and Coolant Circulation - High Water Level B 3.9.4
- BASES LCO (continued)
During this 1 hour period, decay heat is removed by natural circulation to the large mass of water in the refueling cavity. Note 2 allows the required SOC train to be made inoperable for ~ 2 hours per 8 hour period for testing and maintenance provided one SOC train in operation providing flow through the reactor core, and the core outlet temperature is ~ 200°F. The purpose of this Note is to allow the heat flow path from the SOC heat exchanger to be temporarily interrupted for maintenance or testing on the Component Cooling Water or Service Water Systems. During this 2 hour period, the core outlet temperature must be maintained ~ 200°F. Requiring one SOC train to be in operation continues to ensures adequate mixing of the borated coolant. APPLICABILITY One SOC train must be OPERABLE and in operation in MODE 6, with the refueling cavity water level greater than or equal to 647 ft elevation, to provide decay heat removal. The 647 ft elevation was selected because it corresponds to the elevation requirement established for fuel movement in LCO 3.9.6, "Refueling Cavity Water Level." Requirements for the SOC System in other MODES are covered by LCOs in Section 3.4, "Primary Coolant System (PCS)." SOC train requirements in MODE 6, with the refueling cavity water level less than the 647 ft elevation are located in LCO 3.9.5, "Shutdown Cooling (SOC) and Coolant Circulation-Low Water Level." ACTIONS SOC train requirements are met by having one SOC train OPERABLE and in operation, except as permitted in the Note to the LCO~ If one required SOC train is inoperable or not in operation, actions shall be immediately initiated and continued until the SOC train is restored to OPERABLE status and to operation. An immediate Completion Time is necessary for an operator to initiate corrective actions .
- Palisades Nuclear Plant B 3.9.4-3 01/20/98
SOC and Coolant Circulation - High Water Level B 3.9.4
- BASES ACTIONS (continued)
A.2 If SOC train requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can occur through the addition of water with a lower boron concentration than that contained in the PCS. Therefore,, actions that reduce boron concentration shall be suspended immediately. If SOC train requirements are not met, actions shall be taken immediately to suspend loading irradiated fuel assemblies in the core. With no forced circulation cooling, decay heat removal from the core occurs by natural circulation to the heat sink provided by the water above the core. A minimum refueling cavity water level equivalent to the 647 ft elevation provides an adequate available heat sink. Suspending any operation that would increase the decay heat load, such as loading a fuel assembly, is a prudent action under this condition. If SOC train requirements are not met, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed to prevent fission products, if released by a loss of decay heat removal event, from escaping to the environment. The 4 hour Completion Time is based on the low probability of the coolant boiling in that time and allows time for fixing most SOC problems .
- Palisades Nuclear Plant B 3.9.4-4 01/20/98
. SDC and Coolant Circulation - High Water Level B 3.9.4
- 'BASES SURVEILLANCE REQUIREMENTS SR 3.9.4.1 This Surveillance demonstrates that the SDC train is in operation and circulating prima*ry coolant. The flow rate is sufficient to provide decay heat removal capability and to prevent thermal and boron stratification in the core. The 1000 gpm fl ow rate has been determined by operating experience rather than analysis. The Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator in the control room for monitoring the SOC System.
REFERENCES 1. FSAR, Sections 6.1 and 14.3
- Palisades Nuclear Plant B 3.9.4-5 01/20/98
SOC and Coolant Circulation - Low Water Level B 3.9.5
- B 3.9 REFUELING OPERATIONS B 3.9.5 Shutdown Cooling (SOC) and Coolant Circulation - Low Water Level BASES BACKGROUND The purposes of the SOC System in MODE 6 are to remove decay heat and sensible heat from the Primary Coolant System (PCS), as required by the Palisades Nuclear Plant design, to provide mixing of borated coolant, to provide sufficient coolant circulation to minimize the effects of a boron dilution accident, and to prevent boron stratification (Ref. 1). Heat is removed from the PCS by circulating primary coolant through the SOC heat exchanger(s), where the heat is transferred to the Component Cooling Water System via the SOC heat exchanger(s). The coolant is then returned to the PCS via the PCS cold leg(s). Operation of the SOC System for normal cooldown or decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of primary coolant through the SOC heat exchanger(s) and bypassing the heat exchanger(s). Mixing of the primary coolant is maintained
- APPLICABLE by this continuous circulation of reactor primary through the SOC System.
If the primary coolant temperature is not maintained below SAFETY ANALYSES. 200°F, boiling of the primary coolant could result. This could lead to inadequate cooling of the reactor fuel due to the resulting loss of coolant in the reactor vessel. Additionally, boiling of the primary coolant could lead to a reduction in boron concentration in the coolant due.to the boron plating out on components near the areas of the boiling activity, and because of the possible addition of water to the reactor vessel with a lower boron concentration than is required to keep the reactor subcritical. The loss of primary coolant and the reduction of boron concentration in the primary coolant would eventually challenge the integrity of the fuel cladding, which is a fission product barrier. Two trains of the SOC System are required to be OPERABLE, and one train is required to be in operation in MODE 6, with the refueling cavity water level less than the 647 ft elevation to prevent this challenge. SOC and Coolant Circulation-Low Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2)
- Palisades Nuclear Plant B 3.9.5-1 . 01/20/98
SOC and Coolant Circulation - Low Water Level B 3.9.5
- BASES LCO In MODE 6, with the refueling cavity water level less than the 647 ft elevation, both SOC trains must be OPERABLE.
Additionally, one train* of the SOC System must be in operation in order to provide:
- a. Removal of decay heat;
- b. Mixing of borated coolant to minimize the possibility of a criticality; and
- c. Indication of reactor coolant temperature.
An OPERABLE SOC train consists of an SOC pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path and to determine the PCS temperature. The flow path starts in one of the PCS hot legs and is returned to the PCS cold legs. Both SOC pumps may be aligned to the safety injection refueling water tank to support filling the refueling cavity or for performance of required testing .
- APPLICABILITY Two SOC trains are required to be OPERABLE, and one SOC train must be in operation in MODE 6, with the refueling cavity water level less than the 647 ft elevation to provide decay heat removal. Requirements for the.SOC System in other MODES are covered by LCOs in Section 3.4, 11 Primary Coolant System. 11 MODE 6 requirements, with the refueling cavity water level greater than or equal to the 647 ft elevation are covered in LCO 3.9.4, 11 Shutdown Cooling and Coolant Circulation-High Water Level. 11
- Palisades Nuclear Plant B 3.9.5-2 01/20/98
SOC and Coolant Circulation - Low Water Level B 3.9.5
- BASES ACTIONS A.1 and A.2 If one SOC train is inoperable, action shall be immediately initiated and continued until the SDC train is restored to OPERABLE status and to operation, or until a water level of greater than or equal to the 647 ft elevation is established. When the water level is established at the 647 ft elevation or greater, the Applicability will change to that of LCO 3.9.4, Shutdown Cooling and Coolant 11 Circulation - High Water Level," and only one SDC train is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.
If no SDC train is in operation or no SDC trains are OPERABLE, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can occur by the addition of water with lower boron concentration than that contained in the PCS . Therefore, actions that reduce boron concentration shall be suspended immediately. If no SDC train is in operation or no SDC trains are OPERABLE, action shall be initiated immediately and continued without interruption to restore one SDC train to OPERABLE status and operation. Since the plant is in Conditions A and B concurrently, the restoration of two OPERABLE SDC trains and one operating SDC train should be accomplished expeditiously. If no SDC train is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed Immediately. With the SDC train requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded .
- Palisades Nuclear Plant B 3.9.5-3 01/20/98
SOC and Coolant Circulation - Low Water Level B 3.9.5
- BASES SURVEILLANCE REQUIREMENTS SR 3.9.5.1 This Surveillance demonstrates that one SOC train is operating and circulating primary coolant. The flow rate is sufficient to provide decay heat removal capability and to prevent thermal and boron stratification in the core.
In addition, during operation of the SOC train with the water level in the vicinity of the reactor vessel nozzles, the SOC train flow rate determination must also consider the SOC pump suction requirements. The 1000 gpm flow rate has been determined by operating experience rather than analysis. The Frequency of 12 hours is sufficient, considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the SOC System in the control room. SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional SOC pump can be placed in operation, if needed, to maintain decay heat removal and primary coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES 1. FSAR, Sections 6.1 and 14.3
- Palisades Nuclear Plant B 3.9.5-4 01/20/98
Refueling Cavity Water Level B 3.9.6
- B 3.9 REFUELING OPERATIONS B 3.9.6 Refueling Cavity Water Level BASES BACKGROUND The performance of CORE ALTERATION or the movement of irradiated fuel assemblies within containment requires a minimum water level greater than or equal to the 647 ft elevation. During refueling this maintains sufficient water level in the refueling cavity and spent fuel pool.
Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 and 2). Sufficient iodine activity would be retained to limit offsite doses from the accident to less than the guidelines of 10 CFR 100. APPLICABLE During core alterations and during movement of irradiated SAFETY ANALYSES fuel assemblies, the water level in the refueling cavity is an initial condition design parameter in the analysis of the fuel handling accident in containment postulated by Regulatory Guide 1.25 (Ref. 1). A minimum water level of 23 ft (Regulatory Position C.1.c of Ref. 1), which is approximate to an elevation of 647 ft, allows a decontamination factor of 100 (Regulatory Position C.1.g of Ref. 1) to be used in .the accident analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity
.water. The fuel pellet to cladding gap is assumed to contain 12% of the total fuel rod iodine 131 inventory (Ref. 2).
The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level greater than or equal to the 647 ft elevation (and the minimum decay time of 48 hours required by the Operating Requirements Manual) prior to fuel handling, the analysis demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within the guidelines of 10 CFR 100. Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2). . .
- Palisades Nuclear Plant B 3.9.6-1 01/20/98
Refueling Cavity Water Level
. B 3.9.6
- BASES LCO A minimum refueling cavity water level greater than or equal to the 647 ft elevation is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are less than the guideline of 10 CFR 100.
APPLICABILITY LCO 3.9.6 is applicable during CORE ALTERATIONS, and when moving fuel assemblies in the presence of irradiated fuel assemblies in containment. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.14, "Spent Fuel Pool Water Level."
- ACTIONS A.1 and A.2 With a water level below the 647 ft elevation, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies shall be suspended immediately to ensure that a fuel handling accident cannot occur.
The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position. SURVEILLANCE SR 3.9.6.1 REQUIREMENTS Verification of a minimum water level corresponding to the 647 ft elevation ensures that the design basis for the postulated fuel handling accident analysis during refueling operations is met. Water at the required elevation limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2). . The Frequency of 24 hours is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions,
- which make significant unplanned level changes unlikely.
Palisades Nuclear Plant B 3.9.6-2 01/20/98
Refueling Cavity Water Level B 3.9.6
- BASES REFERENCES 1. Regulatory Guide 1.25, March 23, 1972
- 2. FSAR, Section 14.19
- Palisades Nuclear Plant B 3.9.6-3 01/20/98
ATTACHMENT 3 PALISADES NUCLEAR PLANT .
- SECTION 3.9 - REFUELING OPERATIONS CTS'MARKuP AND DISCUSSION
. . - OF CHANGES .- ....
J.~. I
- 3.q I ~11 ~"flt~t-16v - @)
-y.(-~~ REFUELING OPERATIONS Apolicabiljtv
~d"i"\oh> {OJ~ -~
Applies to operating limitations during~ling/opera\'ions) possibility of an ace* ent occurring during coul-d affect public alth and safety.
- 3. a. 1 l--2.E.!!.~~:._---~-_J.:...:_______
The follow ng conditions shall b satisfied during any operation : -J.=efueling
- a. One door oft emergency ir lock shall be pro rly closed.
Whenever bot doors of th personnel air. lock e open during refueling o erations. th equipment door shall be open and the ventilatin syste11 and arcoal _filter in th fuel storage buildin shall be perating. * *
- b. All ~u *atic contain nt isolation valves shall be operable or t least n1 valve in e ch lint shall be clo ed. *
- c. The ontainment ve ing and purge syst s, including two radi ion mo tors that init ate isolation, shal be tested and verifi to bo h be operable mediately prior to efueling operations. The t o monitors sha be located on the ontainment fuel handl ng area evel. (elevatio 649'), shall be p.a of_ the plant area mo itoring syste11 and shal employ ont-out-of- wo logic for isolatio . During normal operat1 n, these monitors 11 not initiate an is lation
*signal. A sw tch*shall -be provi d so that isolation a tion can be initiated du 1n refuelin onl ..
- e. enevtr core g1 try s ting c anged, neutron flu shall be ont1nuously 110 tored by at le t two source range eutron monitors, with each monitor pr iding continuous v sual indication in the contro roOll. When co e geometry is not ing changed, at least one so rc1 range neutr. n monitor shall be n service.
- f. At 1st ont shutdown cool ng pump and heat e hanger shall be *n
~L---~o~p~r~at~i~o~n.:__~----~.../_~----------..L.........-----------'------~
- 3-46 Amendment No. 34 January 27. 1978 I of' 3
- 3. C) _ I
- 3.8 REFUEL!~G OPERATIONS (Continued)
,_;,..:...:.;.:::..:..::~-~~~~--JQ;J)
(sAe) 3.8.2 If any of the conditions in 3.8.1 are not met, all refueling operations Dint.A_,,, shall cease i11111ediately, work shall be initiated to satisfy the required conditions and no operations that may change the reactivity of the core shall b ma e 3.8.3. Refueling operat on shall not be i tiated before t e reactor cor has decayed for a m imum of 48 hours f the reactor h s been operat at ower levels i excess of~ rate 3.8.4 e ventilation ystem and charc l filter in t e fuel storag building shall be operat ng whenever irr iated fuel wh ch has decayed less than 30 days is bei g handled by ei er of the fol owing operatio
- a. Refueli g operation wit the equipment ijoor open, b.
When spent f el which has de yed less than one ear is placed the tilt pit st rage racks, the ulk water tefll1)era re in the tilt it storage ar a must be monito ed continuously t assure that th water temperatu e does not excee lSO"F. Monitori will continue or 24 hour~ af er any addition f fuel to the mai pool or.the ti pit or when a ailure of the sp nt fuel pool cooli g syste111 occur . The equ pment and gen1r1l' rocedures to be il1zed during r fueling are discus ed in th1 FSAR. tailed instructio s, the above sp cifications, and the d sign of th1 fuel ndling equipment ncorporating bu t 1n interlocks 0 and fety features p 1dt assurance th no incident co d occur during the ref ling op1rations at.would result i a hazard to pu ic health and saf. ty.(1) Whenever h~nges are not be ng made in core geometry, one flux mo itor is sufficie . This
- 3-47 Amendment No. 3~. 81 May 22, 1984
- 2. 0 t 3
3.9. l
- 4.2 EQUIPMENT AND SAMPLifffi TESTS TABLE 4.2.1 Minimum Fregyencies for Sampling Tests FSAR Test Section Fregyency REFERENCE
- 1. Reactor Coolant oss Activity Deter- 3 Ti /7 days with a None Samples ination maxim m of 72 hours be-twee samples (T ave gre er than SOO"F).
Gross Ganna by Fission Co tinuous when T ave is None Product Monitor g eater than SOO"Fm. Isotopic analysis /14 days during power for dose equivalent operation 1-131 concentrati~n Radioche11ical for 1/6 months 121 E detenaination Isotopic analysis a) Once/4 hours. whenever for iodine, includin dose equivalent I-131 1-131, 133. 135 exceeds 1.0 ~Ci/gram. and b) Ont sample between Z and 6 hours foll - ing a thenwal po r change exceedin 15' of rated therm& power within a one h ur period.
~hea1~try and 02 ) 3 t1mes/7 days maximu* of 7Z between sampl (T ave greater than lO"F).
(F) Once/30 day and follow-ing lllOd1fi tions or repair to he primary coolant stem involving welding.
- 2. Reactor Cool1nt Boron Concentration Boron Monthly /5<.e)
\_.3. I Boron Co entration Monthly
- See...
c3 ..s) 4-9 Amendinent No. 29, 74, 113. 162 October 26, 1994
- 3.8~
~*
3.'=t.2.. ful(.Q,(' IM'trum\(\-b.*hm REFUELING OPERATIONS Applicability
~---(Applies to operating limitations during refueling operationS')
Objective e possibility of an acci occurring during at could affect public h and safety. The foll wing conditions shall be atisfied during any refuel* g operati ns:
- a. One door f the emergency air loc shall be properly clos .
Wheneve both doors of the person el air lock are open d ing refuel g operations, the equi nt door shall be open d the venti ting syste11 and charcoal filter in the fuel stor ge building shal be operating. .
- b. auto*at1c containment 1s ation valves.shall be perable or at st one valve in each lin shall be closed .
- c. he containment venting 1 purge syste11s, includ ng two radiation monitors that initiate i lation,.shall be*t1st1 and-verified to both be operable illlledi ely prior to refueling operations. The two monitors shall bt cated on the containnie fuel handling area level (elevation 649' , shall *be part of the
- ant area monitoring syste11 and shall etnp y one-out-of~twa logic for isolation. During normal operation, t se monitors will not i 1t1lte an isolation signal. A switch all be provided so tha isolation action can be initiated during fueling only.
f d. Radi~tion 1av11s in the font;linrn!nt) and @:e11t /ue( stotage arMn}--..j-._-~ (s~*3'.,)\ shall bl llOnitored. continuously. . . . cs~~) e. LC.O ng pump an anger sha be in
- 3-46 ;. o-f 5 Amendment No. 34 January 27. 1978
- 3. 9. 'l--
- 3.8 REFUELING OPERATIONS (Continued) en personne in and at all be available in core c.~ At&
3.8.2 If any of the ~onditions in .. are not me, a re ue ing*operations RA.~:l~~ shall cease irrmediately, work shall be initiated to satisfy the required conditions and no operations that may change the reactivity of the core
~*I~ UJJLUJ-WL...IDJUl.5.:..-~~~~~~~~~~~~~~~~~~~_.
3 .8.3. 3.8.4 oa i ter in th fuel storage b lding shall be o erating whenever i adiated fueJ whi has decayed l s than being handled by ither of the foll ing operations:
- a. R fueling operation w h the equipment or
- b. uel handling in th fuel: storage buil ing.
~~- o;
- 3 .8.5. When spent fuel whi has decayed less.than ne year is placed i the tilt pit storage r cks, the* bulk water t erature in the tilt it storage area must be monitored continuous y to assure th~t ~h water temperature does not exceed lSO"F *. Mon1 oring*wn 1 continue or z*4.
hours after an~ additio~ 6f fuel to the ain pool or the ti pit or when a fail ur of the spent fuel ool ool in s stet1 occur e ipment and general pr edures to be utiliz during refueling disc sed in the FSAR. Ott led instructions, t above specificati s, and the esign of the fuel han ing equipment incor rati~g built-in in rlocks an safety features provi assurance that no i cident could occur uring the r ueling operations tha would result in ah ard to public heal and s fety.(l) Whenever ch ges art not being ma e in core geometry, one flux onitor is sufficient. This
- 3-47 2- o-f S Amendment No. 3~. 81 May 22, 1984
3 .S'. <..._
- 3. 17 INSTRUMENTATION SYSTEMS 3.17.6 Action C.otJ D lt~B 3.17.6.l With o e or two Neutron Flux Monitoring channels inoperable:
a) Stop all positive reactivity addttions lnnediately, and b) Be in es, and
- '~)
~ '?*I
.17.6.2 With one CROMs:
a) annel of Rod Position Indication inoperable for o e or more Ve 1fy that the associat rod group is within the li its of S cification 3.10 withi 15 minutes after movement f any rod in t at group~
.17.6.3 With one or two SIRWT Tempt 1tu1"t channels inoperable:
Provide alternate mia s. of t&mperature monitorin 7 days.
.17.6.4 h one Main Feedwater- F ow channels inoperable:
Provide 1-lternate temperatiire channels i nop rab 1e: means of temperature mon* oring within 24 hours. With ont AFW f ow indicator for one or mo e flow path inoperable: a) Oettl"ftlint tht OPERABILITY of tht associ ed AFW flow control valve within 2 hou s.
- flow indicators for one low path inoperable:
a) The assoc ted control valve shal 1 i dhtely be declared inoperabl and the require111ents of .5.2.e apply .
- Amendment Ho. 3 1 67 1 9& 1 98 1 1161 118, 121, 124, !29. 13&, 162 3-74 October 26, 1994 of5 3
- 3.17 INSTRUMENTATION SYSTEMS Table 3.17.6 Instrum1ntation Ooerating Regyirernents foe Other S1fety Fynctjons LCD '? AP~L. Minimum Required OPERABLE Applicable Ia~trum1nt ~biDDll~ ~biDDI]~ ~gcditjgo~
- 1. Neutron Flux Monitoring 2 0 Below io*~ RATED POWER, with fuel in the rea to
- 2. Rod Position 2 1 When more than on CROM is indication capable of rod -Wi hdrawa 1.
- 3. 2'* l Above JOO*F r_ .
- 4. l/Li ne
('e) 0
- 5. I/Lint 0 3~.3 3 .I.(
- 6. 0
- 7.
a. b. 1'ct Air Cool er Condensat l Flow Switch
- 8. rimary Safety Valve I/valve Position Indication PORY 3/valv~* I/valve Abovt z1o*F r_ wh n PORY Position Indication block valve is op n or its position dication syst111 is inope ble.
(a) The provisions (b) The required c anntls shall bt one chan 1 each of 71, 7b, (c) of 71, 7b, Z , or 7d.
- Amendment No. a, i7 1 9&, 98 1 11&1 1181 1211 1241 1291 Hi, 162
. 3* 77 October 26, 1994 L/ ofS
3.9.2-
- 4. 17 INSTRUMENTATION SYSTEMS TESTS Tab 1e 4 . 17.. 6 lnstrymentation Surveillance Reauirements for Other Safety Functions CHANNEL CHANNEL
!nstryment CHECK CALIBRATION SR 3.9.Z. I l Neutron Flux Monitoring 12 hours 18 months ~SeeQl~J.:::77
~ R 3.ll* z. .
,...-~2.~~R~o~d~Po~s~i~t,~*o~n~n"T"~-.-~--.........~~---~
- 3. NA
- 4. Main Not Requ i red
- 5. Main Not Required
- 6. 18 months 7.
fa months
~Se"-
~.I b* 18 months
~.-;
- 18 months .-
3A 1 3.5 Air Cooler Cond1nsa 1 NA 18 months Flow Switch Sa. Primary Safety Valve 18 months acoustical monitor Safety Valve I ?OR~ 12 hours 18 months tailpipe temptratu *-
- J1-1 PORV Acoustical NA-_ -
- 18 *montns*
- 12. hours 18 months
- 10. 12 hours 18 months (a) Once withi 7 days prior to each reacto (b) Vtrificat on of R~ulating Rod Withdra al and Shutdown Rod Inser ion int1rloc s OPERABILITY only, once wit in 92 days prior to each actor startup once prior to startup af reach refueling.
(c) indicator s co11111on to tht safety va ves and PORVs (continued)
- 4-81 s
Amendment No. 6f5
~. 1-44. 171 April S, 1996
3.6 CONTAINMENT SYSTEM Conta'""'.,.f
- fc.11c.'rv"'l.~*1\l e reactivi motion (e ACTION:
With one or mor, containment isolatio valves inoperable (includin6 during perfo nee of valve testing), maintain at least one isolal.ion valve OPERAB in each affected pene ration that is open and eft er:
- a. the inoperable valves to OPERABLE status within 4 ours; or
- b. Iso te each affected penet ation within 4 hours by use f at least on closed and deactivated automatic valve, closed man al valve, or b ind flan e* or
- c. Be i at least HOT SHUTDO N within the next 6 ours and in COLD SHU WN within the foll wing 30 hour$.
3.6.2 The conainment internal pre sure shall not excee :
- a. 1~5 psig when above c 'D SHUTDOWN:and belo HOT STANDBY; and
- b. .0 psig when in POW R OPERATION or HOT s NOBY.
containment intern pressure above the imit, restore press wi in the limit within 1 hour, or be in at ast HOT SHUTDOWN wi hin th *next 6*hours and i ,COLD SHUTDOWN within the following 30 ho rs. 3.6.3
- T. e containment avera e air temperature sh l*not,exceed l40"F hen the lant -is above-COlll, .UTDOWN. With*contai ment average air te perature above the limit, re ore temperature to w'thin the limit with' 8 hours, or be in at least T SHUTDOWN within th next 6 hours and in COLD SHUTDOWN within th following 30 hours.
Two indep.endent c ntainment hydrogen r ombiners shall be O ERABLE when
'the plant i~ in OWER OPERATION.or HOT STANDBY. With one ecombiner inoperable, res re the inoperable re ombiner to OPERABLE tatus within 30 day~ or be i at least HOT SHUTDO within the next 1 hours.
The containme :t purge exhaust and a r room supply isola on valves shall be locked cl ed whenever the plan is above COLD SHUTD WN. With one containment urge exhaust or air om supply isolation valve not locked closed, loc the valve closed wit in 1 hour or be in least HOT STANDBY wi in the next 6 hours nd in COLD SHUTDOWN ithin the following O hours. Entry and exi is permissible throu h a n1ocked* air lo door to perform repairs on o er air lock componen s. Penetration flow paths may be unisolated termittently under a inistrative control 3-40 Amendment No. ~. ~
- I °1 3
- U.o 3.U.b 3 ~
3.9.3 Conto-1"~ Pc-tl~'tiCN --( REFUELING OPERATIONS Applicability
- 3. 8 .1 LLD ~.1'.'!i,C...
I.Lb .rt.s.b. Nell.
..\;
( Sc.!!.3.1.) All autoaat1c containment tso ation valves shall be operable or at ~ I.Lo~~.~ d.1 ll. least one valve in each line shall b1:clos
'i d. Radiation leveh.tn the(conµ'tmnent:>.-and(ipent. ?\lei stofage ar)!as)
- (~~ shall b* 110nttored continuous y.
- e. ts b1tn9 changed, eutron flux shall be v:-* onttnuously 11anttor by at least two sou ce range neutron nitors, wtth each mo itor provtdtng contt uous visual indicat n (See.~~. ii i th* control ro0111. en core geo1111try is ot being changed, a le t one source range utron monitor shall e tn servi
- f. 1east one shutdo be in o eratton.
cec)3 ,,,'f <At.it:> rrs SH9.5.1)-@ (i\nt. LLD '.'>~3. ~ . a.nd Nor~ tor /.c.o ~."\. ~ o..) @
- 3-46 Amendment No. 34 January 27, 1978 2 Ot ~
3.9.:,
- 3.8 REFUELING OPERATIONS (Continued)
- g. During re tor vesse head emoval and whil refueling oper ions are bei performed in th reactor, the r ueling boron conce ration shall be intained in th primary coolant system and shal be checked b s lin on each s 1ft.
- h. room and at 3.8.2 C~oA~ ~~~~~~.;.;.;;..;;;~~~
3.8.3. Refu ling operation. all not bt in iattd bf.fort th reactor core has decay for a minimum f 48 hours if he reactor has ~ en operated at ower evels in excess of ~ rated o r. 3.8.4 ion system and charcoal 1lter 1n the-fuel stor.: ge building shall be perating whenever 1rrad ted fuel which has de yed less than 30 days 1s being handled by eit r of the f.!>11ow.1ng ope tions: Refueling operation wit* th* equipment door op
- Fuel handling in th fuel storage building.
3.8.5. ral procedures to ut111zed during efuel ing are d' cussed in tht FS
- Detailed instr tions, the above pecifications, e design of the u1l handling equip nt incorporating u1lt-in interlo s and safety featur. s provide assuran that no incident ould occur duri g the refueling opera ons that would rt lt in a hazard to public health a safety.(!) Wh ever changes art t being madt in re geometry, on flux monitor is s ficient. This
- 3-47 Amendment No. J~. 81 May 22, 1984 3 of-3
3.9.1
- 2.
b. c. ind Ont or both required SDC trains ..1 bl intentionally rendered 1nop1rabl1 for testing or ..1ntenanc1 for up to 2 ho provided: Pu'B°~* r
~
LC...D* NolC. '"t..
- a. Ont SDC train 1s providing flow through the ructor core,
*anct - _ . _ .
- b.
- Core outlet t-.:>>1t'l'1tun sta s *~ 2oo*F 1n
/ c. Th* refueling cavity water level is ~ 647'~
Act1gn
, l,_ With , ..., OPERABLE .. ins of decay heat r1t10v1l than required:
CotJ D A f\A A.:; * . a. b.
- 2. With less flow through the core than required:
I. Illllldiately suspend all operations involving a reduction f<I\. A-I in PCS boron concentration, and
~A- A:~ Illllld1ately initiate corrective action to return 1 train b.
to op1rat1on.prov1d1ng flow through the core
- i~<(Aooi~m Aalp 3.zsJ Amend111nt No. ~. 173 October 10, 1996
- 3.8 '
~.C,)-{ SDC..ul'\d REFUtLING QPEMTIQHS Appl i cabil i tx l.mLA.r.+ CR.oia.t1af'i
-~lbH W<:..tif" f'nciJ-e L,,t.JJ... 4:--~)
~ W1tri fe..:
10?1._0:lvrft lu:i.W" R~.J1.'- ":t c,47"t:./1.1..1J;.t*t> A9~~ Applies to operating limitations during r fuel "---='-.;....;;..;:~~=..c.i:::~:...,+~ 3 .8 .1
- a. of the emergency a r lock shall be pr perly closed.
Whenev both doors of the ersonnel air lock are open during refuel ng operations, the quipment door sha be open and the venti ating *systet1 and ch rcoal filter in t fuel storage bu' ding shal be operating.
- b. Al auto111t1c contain t isolation valve shall be operabl or at l st one valve in eac line shall be clo td.
- c. e containment vent g and purge syst s, including two adiation nitors that init1 1 isolation, shall be tested and ve ified to both be operable i i1t1ly prior to efueling operat1 ns. The two monitors shall 1 located on the ontain1111nt fuel andl1ng area level (elevation 9'), shall be par of the plantar monitoring system and shall ftlPloy one-out-of-normal operation these monitors wi 1 not initiate logic for 1so1 tion. During isolation signal. A swit h shall* be provide so that isolati n action can be initiated duri refuelin onl *
- d. '-==-Z:..:..:~~::;J* and (~pen§' fue 1. Atcirag1 afeas) -- i 'Ji.
(~_)--r--;;-~u~:a;;;;;-;;;;;~~~TJi;i;r.;;;--;:i;;;~Tn;;~~-;;-tir;rr~---;~(s;f) e. f. Lc.O
- 3-46 Amendment No. 34 January 27, 1978
---~--- - -----
- 3.8 REFUELING OPERATIONS (Continued)
~* Ouri g reactor vessel hea removal and while re ueling operation ar being performed in t reactor, the refuel ng boron c centration shall be intained in the pri *ry coolant syste all be checked by sa lin on each shift.
- h. Dir t co11111unicat on e een personne 1n e con ro room an se.e.\
t refueling machine all be available enever changes inc e ~ ( ~) ornetr are takin 3.8.2 If any of the conditions in 3 . . 1 are not met, a refueling operations shall cease immediately, work shall be initiated to satisfy the required conditions and no aperations that may change the reactivity of the core Refueling opera on shall not bt ini ated before the actor core has decayed for a nimum of 48 hours i the reactor has een operated at power levels *n excess of~ rate o 3.8.4
- b.
If both fan are unavailable, any el 1110vements in p gress shall be comp 1eted d. further fuel move111 s over the spent el storage poo 1 shall be t rminated until one fa 1s returned to se vice. ( S<.t.J 3.( 3.8.5. or ipment-and general 1zed during refuel1n are disc sed in the FSAR. D ailed instructions th* above specifica ions, and the esign of the fuel h dling equipment in rporating built-in nterlocks an safety features pro de assurance that incident could occ during the r ueling operations t t would result in hazard to public he th and fety.(l) Whenever anges are not bting made in cort g1omet , one flux onitor is sufficien . This
- 3-47 Amendment No. 3~. 81 May 22, 1984 3 of L/
- sn c. a.."d ~ C*tt.ik.+1of\ ~U1bN
(@(IPMENT/SMPi4NG.@ TESID-@) WewN-Cfabjlf 4. 2. 2 / (cont i i?6ed l )
£c.utL ~@
-ED
- 12. Iodine Verify the dine Removal System the fo 11 * - * .,g survei 11 an e:
Verify the TSP bask s contain between 8,300 TSP each 18 months Verify that a sa lt from the TSP baskets adjust..nt of b ated water each 18 month . Containment Pu1"91 and Ventilation Isolation alves The ContatnMnt UJ'91 and Ventilation- Isol tion Valves shall dete,..intd clo td:
- a. At 1eas once per 24 hours by chic 1ng the valve pos 1t; n ; nd icator
- 14.
b.. in the control roa11
- At 1 st once every 6 months by performing a leak ra e test between the alves.
- Shutdown Coo1 i ng *
- To ...t the shutdown coonn*g require1111nts 'of Section 3.1.9:
- a. The r1qui Id r11ctor coolant p (s), if not in opera ion should be dettn1in to b4i OPERABLE once tr 7 ctays by v1rHyi g correct breaker a11gnmen~s a~d 1nd1ca power availability
- b. The r u1rtd- st1aa g1n1r1to s) shall be dete,..in d OPERABLE by verif. ing the secondary wa r level ta bt ~-841 t least once per 12 urs.
SR.~.9."1./ c. --At least one coolant loop or train shall bt verified to be in
*f~nc,Y' - operation and circulating reactor coolant at least once per 12
~ hours.
- 15. tn F1tdwat1r Isolation
- a. Verify that the Main F ater Regulating valve a d the associated bypass valve close on n actual or si11uhttd Con ainment High Pressure (CHP) signal once each 18 months.
- b. Verify that the Mai Feedwater Regulating val and the associated bypass valve clos on an actual or simulated ten Generator Low Pressure (SGLP) gnal once each 18 months
- Amendment No. &I-, ~. 1-ii, ~. ~. 165 May 19, 1995 4-13
3.9.5
- 3.J.-9 3 ,q,5 Snc.. d. a ~-t c~~lot'I - L.ow lJ~ W-SHUTQM COOLING CSDCl Spec 1f1cat1 on 3.1.9.3 LCO 1.
z.
.< $'e.e..Q.\Sb 3.+/
or c:ort uy No operations art pt,..1ttld that tht PCS boron co centrat1on or
- z.
Cort outlet t Ont SOC tra (Se_~ \ and
\ 3.'1.4-) b. Cort outl t t111'3tr1tur1 stays ~ zoo*F, *and
- c. Tht ref lin9 cavity water 1 vel ts ~ 647'.
Act ton W1th '""' OPERABLE .. ans of decay heat r1110val th&n required:
- a. Imldhttly tnithtt corrtctht action to nturn a second train to OPERABLE status~ and
- b. . Matnta1n PCS t6')tratur1 as low as practical with Lfl.
avatlablt 1 u1
- z. W1th less f1ow through the core than rtqutrtd:
o Ao. l a. lmldhttly suspend all operations 1nvolv1n9 a reduction
~ ~ tn PCS boron conctntratton, and oA B'2.. b. Imldiattly tntthtt corr1ct1v1 action to return a train
~ '
- to operation providing flow through the core *
@-(Abo Ri&ulf'CJJ ki10.l A* ~ 3 -ZSJ Amend1111nt No. ~. 173 @---< Anb Re.Gui~ Ac:n~iJ 8.~> October 10, 1996 I o 11--
3.9,S
- 4.2 12.
Verify the odine Removal Syst m TSP baskets survei 11 an e: ets contain between ,300 and 11,000 pou. S~e. \ ( ~.s J Verify that a s mi>lt from tht TSP bas tts provides adjust..nt of orated water each 18 nths.
- Containment Purg and Ventilation Isol Th* Conta1n..nt Purge and Ventilation solation Valves dettrm1ntd clo Id:
- a. At ltas once per 24 hours by htck1ng _tht valvt osit1on ind1ca or in th1 ontrol roOll. -
b *.
- 14. Shutdown C~ol incr ..
To ..et tht shutdown cooling requ1remtnts of Section 3.1.9: Tht required rtact9r coolant pump(s), if not in operation should be dtterm1ntd to bt OPERABLE once ptr 7 days by vtr1fy1ng correct breaker al1gnmnts and indicated power availability.
) sha 1 _ be d ani1 ned OPERAB E by level to be -84~ at least nee per
.\
SR :i.9.S. I c. -- At lust Ont coolant loop or r, i sha bt verified to be in PRI&av:..{ operation and circulating reac or coolant at least once per 12 hours. s.
- 1. Y. r1fy that tht Main F dwater R19ul1t1ng valve ypass valve c:lost on n actual or simulated Co Pressure (CHP) signa onct each 18 months.
Verify that tht Ma Feedwater Regulating v vt and the associated bypass valvt clos on an actual or simulat Ste .. Generator Low Pressure (SGLP gnal once each 18 month
- Amendment No. &4-, 9e, Hi, ~. ~. 165 May 19, 1995 4-13
C0-- <At,D llillAI S ik IOhW !>, °i .(,, , "RC.FU!iJ;,;i. 1 C11v1f'f UJ. kr li.tL)
- 3.9 3Jl REFUELING OPEBATIONS 3,Cf,f.o 3,e,,
Su I )
- 3. ~. 2.
~ ."j.3
( :. ."\.Y To minimi e the possibili of an accident ccurring durin 3.~ operatic s that could af ct public healt and safety. 3 .8.1 The f llowing conditi s shall be sati fied during anY, refueling aper tions:
- a. One oor of the emergency air ock shall be roperly closed.
Whe ver both doors of t personnel air lo k are open during ref eling operations, th equipment door s all be open and the ve ilating syste111 and arcoal filter in he fuel storage bui ding sh 11 be operating. S-te.) 3.~.; b. A 1 autoaat1c contain nt isolation val s shall be operabl or at ( 3. .7
~~
east one valvt in a h lint shall bt osed.
- c. The containment ven 1ng and purge sys e11s, including two adiation monitors that init ate isolation, sh 1 be t1st1d and ve ified to both be operable 1 iately prior t refueling operati s. The two monitors shal bt located on th containment fuel ndling area level (elevation 49'), shall be p t of the plant ar monitoring system and shall employ one-out-of two logic for isol tion. During normal operatio , these monitors ill not initiate a isolation signal. A swi ch shall b1 provi d so that isolati action can be initiated dur g refueling only.
- d. Radhtton lev*ls in the [cont~nl!!!nf)~ (spenVfuel storjld'e area~)
sha1=1 bt J10nitoted continuous y. . and heat Se.e. \ ( 3.9..~)
- 3-46 Amendment No. 34 January 27, 1978 I o~ {
- 3.8 Applies to at1ng limitations during refu ling operations.
Objective
.3 .~./ )
( Ste. occurring during 3 ~.t and safety. . l.~.J 1.'*PI 3 .8.1 \ 3.3
- a. One door of the mergency air lock shall be properly losed.
Whenever both ors of the personnel air lock are o en during refueling oper. tions, the equipment door shall be pen and the ventilating stem and charcoal filter in the fue storage building shall be op ating.
- b. All autom 1c containment 1solat1on valves sh 1 be operable or at least on valve in each line shall be closed.
- c. The co ainment venting and purge systems, ncluding two radiation monit s that initiate isolation, shall b tested and verified to both e operable innediately prior to re eling operations. The two onttors shall be located on the co tainment fuel handling area le 1 (elevation 649'), shall be part f the plant area monitoring s tem and shall employ one-out-of-tw logic for isolation. During rmal operation, these monitors wil not initiate an isolation ignal. A switch shall be provide so that isolation action can be initiated during refueling only.
I
..c; cs~~;
- d. Radiation levels in the'fco~tainfftnt)and spent fuel storage shall be monitored continuously.
areas~-+@ 1 R( '
! .. 3 I
- e. core geometry is b ng changed, neutron f x shall be conti ously monitored by least two source ra e neutron
~ moni ors, with each rnonit providing continuou visual indicati n
(~;~iJ in e control room. Wh core geometry ts no being changed, t le st one source ran e utron monitor shall e 1n r *
- f. east one shutdown ooltng pump and h t exchanger shall o ration.
- 3-46 Amendment No. 34 January 27, 1978 I o~ ~
- 3.8 REFUELING OPERATIONS (Continued)
- g. During eactor vesse an w , e re uel ing are ing performed in e reactor, the refueling boro erations con ntration shall be aintained in the primary cool system and sh 1 be checKed b s lin on each shift.
@. ')
I'\ Jo-
~h. Direct conrnunication between personnel in the control room and at the refueling machine shall be available whenever changes in core geometry are taking place.
- 3.8.2 If any of th~ conditions in 3.8.l are not met, all refueling operations~(fJR~I shall cease inrned1ately, .work shall be initiated to satisfy the required f conditions and no operations that may change the reactivity of the core R2..
shall be made. ' Ctb t Refueling operation shall not be initiated before the reactor core has R,, ~ decayed for a minimum of 48 hours if the reactor has been operated at power levels in excess of 2' rated power. 3.8.4 The entilation system and charcoa filter in the fuel storage buil ing sha 1 be operating whenever irrad ated fuel which has decayed les than 3 days is being handled by eith r of the following operations: Refueling operation with he equip111nt door open, or
- 3. 8. 5.
- b. storage building
- If both fans are unavail le, any fuel movetnents in pro completed and further f 1 movements over the spent fue shall be terminated un 1 one fan is returned to erv When spent fuel which has decayed less than one year is placed in the tilt pit storage racks, thi bulk water temperature in the tilt pit storage area must be monitored continuously to assure that the water temperature does not exceed lSO'F. Monitoring will continue for 24 hours after any addition of fuel to the main pool or the tilt pit or when a failure of the spent fuel pool cooling system occurs.
The equipment and general procedures to be utilized during r ueling are discussed in he FSAR. Detailed instructions, the above sp ifications, and the design tht fuel handling equipment incorporating bui t*in interlocks and safety eatures provide assurance that no incident co d occur during the refueling perations that would result in a hazard to pu ic health and safety.( Whenever changes art not being made in core eometry, one flux monitor s sufficient. This s~~ ( i.9.I 3.1.'Z-J .").3
- J.V-1 3.47 Amendment No. 3~. 81 May 22, 1984
- 4.2 EQUIPMENT ANO SAMPLING TESTS Table 4.2.1 (continued)
Minimum Ereauencies for Samol1ng Tests f SAR Section Test Ereayency REFERENCE B ron Concentration Monthlym 9.4 Bulk Water Temperature Continuously when None bundles are stored in tilt pit racks with less than one year deca ,.,
- 7. Secondary Coolant Coolant Gross Radi
- 3 times/7 days wi None activ1ty a maximu* of 72 between samples Isotop1c Analys a) 1 per 31 da~ ,
Dose Equivalen whenever th Concentration gross act i 1ty detenninat n indicates iodine concentr 1ons (~~) greater han 10!. of the llowable limit b) 1 per. 6 months, when ver the gross act ity de nnination in 1cates iodine c ncentrat1ons 1ow lei of tht l lowable l111it
- 1) tor is (2) After at east 2 EFPO and at least 20 days since the last shutdo (s~~)
3, "{ than 48 ours.
@: ;? s) oe!ete(J)-@
- 6) Reference Specification 3.8.5 for maximum bulk water temperature and monitorin re uirements.
- 7) Refer nee Basts section of Specificat n 3.8 and Section S.4.2f of. the Design Fea res for minimum boron concentra on (~ 1720 ppm) .
Amendment Ho. 291 11Q 1 123, 162 October 26, 1994 4*10 3 of Lj
4.2 Table 4.2.2
- 1. CONTROL RODS for E ui ment Tests Refueling FSAR Section REFERENCE 7.6.1.3 2* CONTROL RODS Parti l Movement 7.6.1.3 of al Rods (Mini um of 6 In)
~
- 3. Pressurizer One Each 4.3.7 Safety Valves Refueling (5.,-e. )-- 4. Main Steam Five Each 4.3.4
~-'*1 Safety Valves Refueling
~- ( 5. Refueling System Functioning Prior to 9.11.4
~ Interlocks Refueling Operations
- 6. Service Water unctioning 9.1.2 System Valve Actuation on SIS and RAS
- 7. Primary System 4.7.1 Leakage
- 8. Deleted
- 9. Boric Acid Verify proper Heat Tracing temperature readings.
- 10. Safety lnjec on Verify that level and Tank Level ad. pressure indication Pressure is between independent high high/low alarms for level and pressure.
Amendment No. Ho, ~' m, ~' +ii, ~' ~'
"ELEC CHANGES" 4-11
- ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.1, BORON CONCENTRATION ADMINISTRATIVE CHANGES (A)
A.1 All reformatting and renumbering are in accordance with.NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more_ details does not result in a technical change. A.2 The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases. A.3 CTS 3.8.2 provides the required actions if the refueling boron concentration in the primary coolant system is not maintained. CTS 3.8.2 establishes measures that are considered equivalent to the Required Actions of ITS 3. 9 .1, Condition A. Although the exact wording of the CTS and ITS are not the same, the intent of both the CTS and ITS is to immediately suspend any activity which may result in the addition of positive reactivity, and to immediately initiate actions to restore the boron concentration within the specified limit. Therefore, the revised wording of CTS 3.8.2 is considered editorial in nature .
- Palisades Nuclear Plant Page 1of3 01720/98
- MORE RESTRICTIVE CHANGES (M)
ATTACI'ENT 3 DISCUSSION O:t: CHANGES SPECIFICATION 3.9.1, BORON CONCENTRATION M.1 CTS 3.8.lg establishes the requirement for the minimum boron concentration "during reactor vessel head removal and while refueling operations are being performed in the reactor." The CTS defines "refueling operations" as "any operation involving movement of core components (except for incore detectors) when the reactor vessel head is untensioned or removed with fuel in the reactor vessel. " In the ITS, the comparable requirement for boron concentration is met in proposed ITS 3. 9 .1, "Boron Concentration." The Applicability of proposed ITS 3.9.1 is MODE 6. MODE 6 is entered with the detensioning of the first reactor vessel head stud and remains in effect as long as fuel is in the vessel or the reactor vessel head is retensioned. As such, the Applicability of ITS 3.9.1 is more restrictive than CTS 3.8. lg since it addresses the entire period the reactor vessel head is detensioning with fuel in the reactor vessel inclusive of reactor vessel head removal and refueling operations as defined in the CTS. Therefore, this change is an additional restriction on plant operations and is consistent with NUREG-1432. M.2 CTS 3.8.lg requires the *Primary Coolant System be maintained at the "refueling boron concentration. " Proposed ITS 3. 9 .1 requires the boron concentration of the primary coolant system and the refueling cavity be maintained at the "Refueling Boron Concentration." This additional restriction helps ensure that the single mass (water volume) formed by the primary coolant system and the refueling cavity will maintain the reactor subcritical while.in MODE 6. This change is an additional restriction on plant operations and is consistent with NUREG-1432. M.3 CTS 4.2, Table 4.2.1 item 2 specifies that the boron concentration of the reactor
- coolant (primary coolant system) be tested "twice per week." Therefore, the requirement of the CTS is met if the reactor coolant is tested every 84 hours whenever the reactor vessel head is removed and fuel is contained in the reactor vessel. (Note:
the sampling frequency during reactor vessel head removal and refuelirig operations is addressed by CTS 3.8.lg.). In proposed ITS 3.9.1 for this same plant condition (i.e., MODE 6 ), the boron concentration of the primary coolant system and the refueling cavity must be verified to be within limit every 72 hours. As such, the ITS imposes a more restrictive sampling frequency and a new sampling location (refueling cavity) than the CTS. This change is an additional restriction on plant operations and is consistent with NUREG-1432 .
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- ATTAC1'ENT 3 DISCUSSION OE CHANGES SPECIFICATION 3.9.1, BORON CONCENTRATION LESS RESTRICTIVE CHANGES - REMOVAL .OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
There were no "Removal of Detail" changes associated with this specification. LESS RESTRICTIVE CHANGES (L) L.1 The Frequency for verifying boron concentration in CTS 3.8.lg has been changed from "on each shift" to "72 hours." Considering the large volume of water in the primary. coolant system (and refueling cavity during Core Alterations), in addition to administrative controls instituted to preclude a boron dilution event, a sampling Frequency of 72 hours is adequate to identify slow changes in boron concentration. Furthermore, if a rapid change in boron concentration would occur during refueling operations, an increase in subcritical multiplication would be detected by the source range nuclear instrumentation required by proposed LCO 3.9.2, "Nuclear Instrumentation." Therefore, based on the availability of the source range nuclear
- instrumentation, the large volume of water in the primary coolant system (and refueling cavity during Core Alteration), and administrative control instituted to preclude a boron dilution during refueling operations, performing a verification of the boron concentration on a 72 hour Frequency is considered acceptable. This change is consistent with NUREG-1432 .
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- . ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3. DISCUSSION OF CHANGES SPECIFICATION 3.9.2, NUCLEAR INSTRuMENTATION A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Tecliiiical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or. English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been. added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 The Bases of the current Technical Specifications for this section have been completely
- A.3 replaced by revised Bases that reflect the format and applicable content consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases.
CTS 3.8.le establishes the requirements for neutron flux instrumentation during refueling operations. The CTS defines "refueling operations" as "any operation involving movement of core components (except for incore detectors) when the reactor vessel head is untensioned or removed with fuel in the reactor vessel." CTS 3.8. le requires at least two source range _neutron monitors whenever core geometry is being changed, and one source range neutron monitor when core geometry is not being changed. In addition to CTS 3.8. le, two channels of neutron flux monitoring instrumentation are required by CTS Table 3.17.6item1 whenever the plant is below 10-4 % Rated Power with fuel in the reactor. For plant conditions which exist. during refueling operations (i.e., the boron concentration of the primary coolant system is at the Refueling Boron Concentration) the neutron flux monitoring requirement of CTS Table 3.17.6 item 1 is fulfilled by the source range instruments. As such, the requirements ofCTS 3.17.6item1 bound the requirements of CTS 3.8.le since the applicability of CTS 3.17.6 item 1(below104 % Rated Power with fuel in the reactor) includes the applicability of CTS 3. 8 .1 e (during refueling operations);
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ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION A.3 (continued) This means that although CTS 3.8.le only requires one source range neutron monitor to. be in-service when core geometry is not being changed; two neutron flux monitors are still required to be Operable by CTS 3 .17. 6 item 1. In proposed ITS 3. 9. 2, two source range channels are required to be Operable whenever the plant is in MODE* 6.
- Therefore, the requirement of CTS 3.8.le and CTS 3.17.6 item 1 are equivalent to the requirements of ITS 3.9.2.
A.4 CTS 3.8.2 provides the required actions if less than the required number of neutron flux monitors are available. CTS 3.8.2 establishes measures that are considered equivalent to ITS 3.9.2 Required Actions A.1, A.2 and B.l. Although the exact wording of the CTS and ITS are not the same, the intent of both the CTS and ITS is to immediately suspend any activity which may result in the addition of positive reactivity, and to immediately initiate actions to restore the required number of neutron flux monitors to an Operable status. Therefore, the revised wording of CTS
- 3. 8. 2 is considered editorial in nature.
- .MORE RESTRICTIVE CHANGES (M)
There were no "More Restrictive" changes associated with this specification. LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.1 CTS 3.8.le contains specific details regarding the requirements for indication of the source range monitors, i.e., "with each monitor providing continuous visual indication in the control room." These details describe elements of the source range instruments which are addressed by the definition of OPERABILITY and are not directly pertinent to the actual requirement, i.e., Limiting Condition for Operation. Since these details are not necessary to adequately describe the actual regulatory requrrement, they can be moved to a licensee controlled document without a significant impact on safety. Placing these details in the Bases of ITS 3.9.2 provides adequate assurance that they will be maintained. The Bases are controlled by the Bases Control Process in Chapter 5 of the proposed Technical Specifications. This change is consistent with NUREG-1432 .
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- LESS RESTRICTIVE CHANGES (L)
ATTACHMENT 3 DISCUSSION OF CHANGES . SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION L.1 CTS 3 .17 .6 requires two channels of Neutron Flux Monitoring to be Operable below 104 % Rated Power. CTS 3.17.6.lc requires that shutdown margin be verified within 4 hours, and once each 12 thereafter, whenever one or two Neutron Flux Monitoring channels are inoperable. Proposed ITS 3.9.2 Required Action B.2 only requires a boron concentration verification (which ensures an adequate shutdown margin for existing core conditions) if two source range channels are inoperable, and only requires the verification to be performed once per 12 hours. Both the CTS and ITS require two Operable source range channels to ensure redundant monitoring capability is available to detect changes in core conditions. With one channel inoperable redundancy will be lost, however, one channel is still available to provide direct means for monitoring core reactivity. Since the capability exists to directly monitor core reactivity conditions with one source range channel inoperable, and core reactivity conditions are stabilized due to the suspension of Core Alterations and positive reactivity additions, there is no need to verify core reactivity conditions by use of chemical analysis other than at the normal sampling frequency of 72 hours. In addition, the accelerated initial performance (within 4 hours) of the boron concentration verification when two source range channels are inoperable is excessively restrictive and not warranted. This is based on routine sampling (every 72 hours) and knowledge of stable conditions prior to the loss of the source range channel, and the recognition that a PCS dilution event is detectable through other means such as an uncontrolled increase in the refueling cavity level. - This change is consistent with NUREG-1432 as modified by TSTF-96 .
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- ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.3, CONTAINMENT PE~ETRATIONS A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformattfug, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording
. preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.
- A.2 The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content
- A.3 consistent with NUREG-1432. The revised Bases are shown in the proposed Technical Specification Bases.
The requirements of CTS 3.8.la, CTS 3.8.lb and CTS 3.8.lc address containment air locks, automatic containment isolation valves, and the containment vent and purge system respectively, during refueling operations. The CTS defines "refueling operations" as "any operation involving movement of core components (except for incore detectors) when the reactor vessel head is untensioned or removed with fuel in the reactor vessel. " In the ITS, comparable requirements are met in proposed ITS 3.9.3, Containment Penetrations. The Applicability of ITS 3.9.3 is stated as "during Core Alterations" and "during movement of irradiated fuel assemblies within containment. " Core Alterations is defined, in part, "as the movement or manipulation of any fuel, sources, or reactivity control component within the reactor vessel with the vessel head removed and fuel in the vessel. " Although the wording of the CTS applicability and ITS Applicability are not exact, the intent of both the CTS and ITS is to address the condition where a potential fuel handling accident. As such, the minor change is wording is considered to be editorial in nature .
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- A.4 SPECIFICATION 3.9.3, CONTAINMENT PEN"ETRATIONS ATTACHMENT 3 DISCUSSION OF CHANGES In CTS 3.8. la, the requirement that "one door in the personnel air lock be closed" has been added for clarity. This requirement is implicit from the CTS requirement that one door of the emergency air lock shall be properly closed, and the statement that "whenever both doors of the personnel air lock are open during refueling operations, the equipment hatch shall be open and the ventilation system and charcoal filter in the fuel storage building shall be operating. " The addition of this requirement does not change the original intent of the CTS requirement but simply clarifies the way the LCO can be met. Therefore, this change is considered administrative in nature.
- A.5 CTS 3.8.2 provides the required actions if the containment air lock doors or containment isolation valves are not in their required position. CTS 3.8.2 establishes measures that are considered equivalent to the Required Actions of ITS 3. 9. 3, Condition A. Although the exact wording of the CTS and ITS are not the same, the intent of both the CTS and ITS is to imm_ediately suspend any activity which may result in a fuel handling accident (e.g., Core Alterations and movement of irradiate fuel assemblies). Therefore, the revised wording of CTS 3.8.2 is considered editorial in nature.
- A.6 CTS 3.8.2 stipulates that work shall be initiated to satisfy the condition and no operations that may change the reactivity of the core shall be made if the containment air lock doors or containment isolation valves are not in their required position. The inclusion of this statement in the CTS or ITS is not necessary since the plant is removed from the mode of applicability (i.e., Refueling Operations for the CTS, during Core Alterations and movement of irradiated fuel assemblies for the ITS) as soon as the refueling activities are. suspended. As such, the elimination of this statement is considered administrative in nature since restoration of the inoperable component and subsequent compliance with the LCO is only required when the plant is in the specified Mode of Applicability .
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- A.7 SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS ATTAC1'ENT 3 DISCUSSION OE CHANGES CTS 3. 8 .1 has been modified to include an explicit requirement for the containment equipment hatch. Proposed LCO 3.9.3a requires the equipment hatch to be closed and held in place by four bolts. LCO 3.9.3a is also modified by a Note which states that this requirement is only required when the fuel handling building ventilation system is not in compliance with (proposed) LCO 3.7.12. The option to maintain the equipment.**
hatch closed during refueling operations is implicitly required in CTS 3. 8 .4 which requires the fuel storage (handling) building ventilation system be in operation when the equipment hatch open. Conversely, if the equipment hatch is closed, then the fuel
- storage ventilation system is not required to be in operation. As such, this proposed change is categorized as administrative since it does not impose any additional requirements, but simply establishes a convention similar to NUREG-1432 while maintaining the allowance for equipment hatch operation presently contained in the CTS.
Specifying that the equipment hatch be held closed by four bolts is based on good engineering practice and has been adopted from NUREG-1432 .
- MORE RESTRICTIVE CHANGES (M)
M.1 CTS 3.6. lb states that containment integrity shall not be violated when the reactor vessel head is removed (unless the PCS boron concentration is at the Refueling Boron Concentration). Furthermore, the actions associated with CTS 3.6.1 only address the condition of one or more containment isolation valve(s) inoperable (they do not address other penetrations such as containment air locks), and allow 4 hours to either restore the inoperable valve(s) to operable status or, to isolate the affected penetration. As such, the CTS does not require containment integrity when the reactor vessel head is removed as long as the requirement for boron concentration is satisfied and, if containment integrity is required (due to a low boron concentration) the CTS allows 4 hours to re-establish containment integrity as a result of an inoperable valve. In the ITS, the plant condition which exists when the reactor vessel head is removed is defined as MODE 6. While in MODE 6, ITS 3.9.1, "Boron Concentration" requires the boron concentration in the primary coolant system and the refueling cavity to be at the Refueling Boron Concentration. If the boron concentration is not within the limit, actions are immediately initiated to restore the required boron concentration.
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- M.1 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS Unlike the requirements of the CTS, the ITS does not permit the option to maintain containnient integrity in lieu of establishing the required boron concentration while the reactor vessel head is removed. This change is an additional restriction on plant operations and is consistent with NUREG -1432. M.2 A new SR has been adopted to address containment penetrations during Core Alterations and movement of irradiated fuel assemblies within containment. Proposed SR 3. 9. 3 .1 requires a verification that each required containment penetration is in the required position and is verified on a Frequency of every 7 days. This surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The surveillance on valves in unisolated penetrations* will demonstrate that the valves are not blocked from closing and that each valve operator has motive power to ensure the valve is capable of closing on an actuation signal. This change is an additional restriction on plant operations and is consistent with NUREG-1432 .
- LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
There were no "Removal of Detail" changes associated with this specification. LESS RESTRICTIVE CHANGES (L) L.l CTS 3.8.lb requires all automatic containment isolation valves be operable or that at least 0ne valve in each line be closed. Proposed ITS 3.9.3d requires each penetration providing direct access from the containment atmosphere to th~ outside atmosphere be
- either: 1) closed by a manual valve, automatic isolation valve, blind flange, or equivalent, or; 2) capable of being closed by an Operable Containment Refueling Radiation Monitor. The requirements for containment penetration closure associated with ITS 3. 9. 3 are relaxed from the requirements of CTS 3. 8 .1 b since they only address components in systems which provide a direct path from the containment atmosphere to the outside atmosphere .
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- L. l (continued)
ATTACHMENT 3 DISCUSSION OE CHANGES SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS The reduEtion in the scope of penetrations from the CTS to the ITS is acceptable since the potential for containment pressurization as a result of an accident is not likely while
- in MODE 6. Without containment pressurization, only those penetrations which provide a direct path from the containment atmosphere to the outside atmosphere present a potential release path for fission product radioactivity. The less stringent requirements of ITS 3.9.3 continue to ensure that a release of fission product radioactivity within containment as a result of a fuel handling accident will b~ restricted from escaping to. the environment. This change is consistent with NUREG-1432.
L.2 CTS 3.8. lc requires the containment venting and purge systems be tested and verified Operable "immediately prior to refueling operations." Proposed ITS 3.9.3 and SR 3. 0 .4 require the Operability of each containment purge and exhaust valves prior to entering the mode of Applicability ( e.g., prior to Core Alterations or the movement of irradiated fuel assemblies within the containment) .. The CTS has been revised to delete the "immediately prior to _refueling operation" requirement. This is .acceptable since
- the ITS provides general rules for the application of surveillance requirements in the technical specifications. SR 3.0.4 establishes the requirement that applicable SRs must be met before entry into a mode or other specified condition in the Applicability. In addition, the specific time frames and conditions necessary for meeting the SRs are specified in the Frequency. Although the phase "immediately prior to refueling operations" implies a conditional type frequency, proposed SR 3.9.3.2 specifies a fixed Frequency of 18 months. The 18 month Frequency maintains consistency with other similar instrumentation (e.g., ESFAS) and valve testing requirements and is considered acceptable since the containment venting and purge systems also requires the performance of additional SRs to ensure Operability of the isolation function. The SRs for the actuating instrumentation are contained in proposed ITS 3.3.6, "Refueling CHR Instrumentation" which also has an Applicability of during Core Alterations and during movement of irradiated fuel assemblies in containment. Therefore, SR 3.9.3.2 is only a system functional test of the valves to close on a signal while the actuation signal are addressed in ITS 3.3.6. This change is consistent with NUREG-1432. *
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- ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL ADMINISTRATIVE CHANGES (A)
A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventiOns were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change.
- A.2 The Bases of the current Technical Specifications for this section have been completely replaced by revised Bases that reflect the format and applicable content
- A. 3 consistent with NUREG-1432. The*revised Bases are shown in the proposed Technical Specification Bases.
CTS 3 .1. 9. 3 contains requirements for shutdown cooling (SDC) when there is fuel in the reactor and the primary coolant system temperature is < 200
- F and the primary
- coolant system loops are not filled. The CTS specifies, in part, that one SDC train shall be in operation and at least two means of decay heat removal shall be Operable.
The available means of decay heat removal are; SDC train A, SDC train B or, the refueling cavity with a water level ~ the 647' elevation. Although not explicitly stated, if one of the means for decay heat removal* is the refueling cavity, then the plant must be in a refueling configuration since the 647' elevation is approximately 2' below the top of the refueling cavity. This plant condition is comparable to the plant condition addressed in proposed ITS 3. 9 .4, "Shutdown Cooling and Coolant Circulation - High Water Level. " ITS 3. 9 .4 requires one SDC train to be Operable and in operation. The Applicability of ITS 3.9.4 is MODE 6 with the refueling cavity water level ~647' elevation. As such, when the refueling cavity water level is relied upon as a means for decay heat removal, then the requirements of CTS 3.1.9.3 and ITS 3. 9 .4 are the same .
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- SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL A.3 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES When the refueling cavity water level is NOT relied upon as a means of decay heat removar(i.e., water level <647' elevation), then the requirements of CTS 3.1.9.3 are addressed by proposed ITS 3. 9. 5, "SDC and Coolant Circulation - Low Water Level." Therefore, the changes associated with CTS 3 .1. 9. 3 are administrative in nature since they do not modify the requirements of the CTS. A.4 CTS 3.8. lf establishes the requirement for a shutdown cooling pump during refueling operations and stipulates that at least one shutdown cooling pump be in operation. The requirement of CTS 3. 8 .1 f has been modified to require one SDC train Operable in addition to one SDC pump in operation in order to maintain consistency with proposed ITS 3.9.4. This change has been characterized as administrative since the
- requirements of CTS 3. 8. lf are encompassed by the requirements of CTS 3 .1. 9. 3.
CTS 3 .1. 9. 3 requires a SDC train to be Operable and in operation when the refueling cavity water level is :<: 647' elevation. CTS 3 .1. 9. 3 applies whenever there is fuel in the reactor with the primary coolant. system temperature .< 200
- F and the primary coolant loops not filled. As such, CTS 3. 8. lf and CTS 3 .1. 9. 3 establish requirements that are equivalent to those specified in ITS 3.9.4.
A.5 CTS 3.8.2 provides the required actions when one shutdown cooling pump is not in operation. Proposed ITS 3.9.4 Condition A addresses that same plant condition and provides the appropriate Required Actions (Required Actions A. l , A. 2, and A. 3). Although the exact wording of the CTS 3.8.2 and ITS 3.9.4 are not the same, the intent of both the CTS and ITS is to immediately suspend any activity which may result in the addition of positive reactivity, immediately suspend loading irradiated fuel assemblies in the core, and to immediately initiate actions to restore compliance with the LCO. Therefore, the revised wording of CTS 3.8.2 is considered editorial in nature .
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- ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WAfER LEVEL MORE RESTRICTIVE CHANGES (M)
M .1 CTS 3 .1. 9. 3 Exceptions 1 and 2 provide allowances to suspend all SDC flow through the reactor core (exception 1) and to render the required SDC trains inoperable for testing or maintenance (exception 2). Proposed ITS 3. 9 .4 also contains these allowances (LCO Note 1 and 2) but restricts their use in any 8 hour period. The , intent of this change is to prescribe limits on the frequency these exceptions may be utilized and to avoid the potential misapplication of their use by repeatedly relying on the exception. Although the 8 hour period has no analytically basis, it has been included in the ITS to maintain consistency with NUREG-1432. As such, this is an additional restriction on plant operations. M.2 A new Requir~d Action has been added to CTS 3.1.9.3 and CTS 3.8.2 to address the condition when the requirements for a SDC train are not met. The proposed change requires all containment penetration providing direct access from the containment atmosphere to the outside atmosphere be closed within four hours. The intent of this Required Action is to prevent radioactive fission products from escaping the containment building if released by a loss of decay heat event. The four hour
- Completion Time is a reasonable time to fix most SDC problems and is based on the low probability of the primary coolant boiling in that time. This change is an additional restriction on plant operations and is consistent with NUREG-1432.
LESS RESTRICTIVE.CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA) LA.1 In CTS 3 .1. 9. 3, the details associated with SDC train Operability have been moved to the Bases of proposed ITS 3.9.4. The CTS states that an Operable SDC train consists of "an Operable SDC pump and an Operable SDC heat flow path to Lake Michigan." In the ITS, the details of what constitutes an Operable SDC train are contained in the Bases. As such, .the reference to the SDC pumps and heat flow paths in CTS 3. 1. 9. 3 have been moved to the Bases. This change is acceptable since this information provides details of design which are not directly pertinent to the actual requirement. Since these details are not necessary to adequately describe actual regulatory requirements, they can-be moved to a license controlled document without a significant impact on safety. Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are controlled by the Bases ControlProgram in proposed ITS Chapter 5.0. *
- Palisades Nuclear Plant Page3 of 5 01/20/98
- SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL ATTACHMENT 3 DISCUSSION OF CHANGES LA.2 In CTS 3.1.9.3 when there are fewer Operable means of decay heat removal than required, Action 1. b states that the primary coolant system temperature should be maintained as low as practical with available equipment. In ITS 3.9.4, a comparable condition exists when SDC train loop requirements are not met. However, ITS 3. 9 .4 does not contain explicit instructions to maintain the primary coolant system as low as practical with available equipment since this action is beyond the scope of the LCO (i.e., restore compliance with the LCO). Off Normal procedures are used to address alternate ways to maintain the primary coolant system temperature. as low as practical when a loss of shutdown cooling exist. As such, CTS Action 1. b has been removed from the CTS and placed in plant procedures. This change is acceptable since these details are not necessary to adequately describe the actual regulatory requirement and placing this information in license controlled documents will not result in a.significant impact on safety. This change is consistent with NUREG-1432.
LA.3 CTS 3.8.lf specifies, in part, that one (SDC) heat exchanger shall be in operation. ITS 3. 9 .4 specifies that one SDC train shall be Operable and in operation. In the ITS, the details of what constitutes an Operable SDC train are contained in the Bases. As such, the reference to the heat exchangers in CTS 3. 8 .1 f has been moved to the Bases. This change is acceptable since this information provides details of design which are not directly pertinent to the actual requirement. Since these details are not necessary to adequately describe actual regulatory requirements, they can be moved to a license controlled document without a significant impact on safety. Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are controlled by the Bases Control Program in proposed ITS Chapter 5.0 .
- Palisades Nuclear Plant Page 4 of 5 01/20/98
ATTACHMENT 3
. . DISCUSSION OF CHANGES SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL LESS RESTRICTIVE CHANGES (L)
L.1 CTS 3. 1. 9. 3 allows all flow through the reactor core to be intentionally stopped for up to 1 hour provided, in part, that the core outlet temperature stays ~ 200
- F and two SDC trains are Operable. Proposed ITS 3.9.4 does not contain these additional ...
restrictions. While in MODE 6 with the refueling cavity water level :<: 647' elevation, an increase in primary coolant. system temperature above 200
- F is not an immediate concern. The affects of elevated coolant temperatures at or above. the boiling point would eventually challenge the integrity of the fuel cladding, which is a fission product barrier, and lead to a reduction in boron concentration due to boron plating-out on components near the area of boiling. However, due to the relative short time flow is allowed to be suspended (up to 1 hour per 8 hour period), sufficient
. boiling would not occur such that it would result in a signification reduction in the boron concentration or present a challenge to the fission product barrier. Coolant temperatures above the saturation temperature with no forced circulation become an immediate concern only when the reactor vessel head is installed due to the potential of vapor formations in the primary coolant system loops. The additional restriction in the CTS to maintain two SDC trains Operable when all flow through the reactor core is intentionally stopped is excessively restrictive since two redundant heat removal methods are still available. That is, when flow is stopped, one SDC train is still required to be Operable and the refueling cavity water level is still required to be
- <: 647' elevation thus providing adequate and redundant heat removal capability.
This change is consistent with NDREG-1432 .
- Palisades Nuclear Plant Page 5 of 5 01/20/98 .
- SPECIFICATION 3.9.5 SDC & COOLANT CIRCULATION - LOW WATER LEVEL ADMINISTRATIVE CHANGES (A)
ATTACHMENT 3 DISCUSSION OF CHANGES A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involves no technical changes to existing Technical Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS. Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A. 2 CTS 3 .1. 9. 3 specifies the requirements for SDC trains and is applicable whenever there is fuel in the reactor with the primary coolant system temperature < 200
- F and
- the primary coolant system loops not filled. Proposed ITS 3. 9. 5, "SDC and Coolant Circulation -Low Water Level" also specifies the requirements for SDC trains and is applicable in MODE 6 with the water level less than the 647' elevation. The applicability of CTS 3 .1. 9. 3 encompasses the applicability of ITS 3. 9. 5 since the CTS essentially addresses a plant condition equivalent to MODE 5 in the ITS (cold shutdown) and remains in affect as long as fuel is in the reactor. The structure of the ITS is such that it contains a separate specifications to address SDC train requirements in MODE 5 and MODE 6. As such, the inclusion of ITS 3.9.5 Applicability in CTS 3 .1. 9. 3 is considered administrative since it does not result in a change from the CTS requirements and is provided to support presentation of the ITS structure only .
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- SPECIFICATION 3.9.5 SDC & COOLANT CIRCULATION - LOW WATER LEVEL A.3 ATTACHMENT 3 DISCUSSION OF CHANGES CTS 3.1.9.3 Action 1 specifies the required actions when there are fewer Operable means of decay heat removal than required. The CTS actions state to "immediately initiate corrective actions to return a second train to Operable status." The LCO of CTS 3. l-; 9. 3 lists SDC train A, SDC train B and the refueling cavity with a water level ~ 647' as the available means of decay heat removal. Therefore, restoration of.
either SDC train, or establishing a refueling cavity water level ~ 647' would restore compliance with the LCO. In proposed ITS 3.9.5, whenever an SDC train is inoperable, the Required Actions allow the option to either restore an SDC train to Operable status, or to initiate action to establish a refueling cavity water level ~ 647' elevation. Although the presentation of the Required Actions in the ITS is different from that in the CTS, the intent of the ITS and CTS are the same since both documents prescribe similar methods to restore compliance with the LCO. MORE RESTRICTIVE CHANGES (M) M.l A new Required Action has been added to CTS 3.1.9.3 and CTS 3.8.2 to address the condition when no SDC trains are Operable or in operation. The proposed change requires all containment penetration providing direct access from the containment atmosphere to the outside atmosphere be closed within 4 hours. The intent of this Required Action is to prevent radioactive fission products from escaping the containment building if released by a loss of decay heat event. The 4 hour Completion Time is a reasonable time to fix most SDC problems and is based on the low probability of the primary coolant boiling in that time. This change is an additional restriction on plant operations and is consistent with NUREG-1432. M.2 CTS 3.1.9.3 contains an exception for SDC train operability when there is fuel in the reactor with PCS temperature < 200°F and the PCS loops are not filled. The exception allows all flow through the reactor core to be intentionally stopped for up to 1 hour. The CTS does not restrict the use of this exception based on level in the refueling cavity. Therefore, all SDC flow is permitted to be stopped for up to 1 hour when the refueling cavity water level is < 647 ft elevation. In proposed ITS 3. 9. 5 (refueling cavity water level < 647 ft elevation), the allowance to stop all SDC flow is not permitted since there is no assurance that an adequate mass of water is available to provide the necessary decay heat sink when the refueling cavity is < 647 ft elevation. As such, the deletion of this allowance in the ITS is more restrictive than currently allowed in the CTS. This change is consistent with NUREG-1432 .
- Palisades Nuclear Plant Page 2 of 3 01/20/98
- ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.5 SDC & COOLANT CIRCULATION - LOW WATER LEVEL LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 In CTS 3.1.9.3, the details associated with SDC train Operability have been moved to the Bases of proposed ITS 3. 9. 5. The CTS states that an Operable SDC train consist of "an Operable SDC pump and an Operable SDC heat flow path to Lake Michigan." In the ITS, the details of what constitutes an Operable SDC train are contained in the Bases. As such, the reference to the SDC pumps and heat flow paths in CTS 3.1.9.3 have been moved to the Bases. This change is acceptable since this information provides details of design which are not directly pertinent to the actual requirement. Since these details are not necessary to adequately describe actual regulatory requirements, they can be moved to a license controlled document without a significant impact on safety. Placing these details in the Bases provides adequate assurance that they will be maintained since the Bases are controlled by the Bases Control Program in proposed ITS Chapter 5.0. LA.2 In CTS 3.1.9.3 when there is fewer Operable means of decay heat removal than required, Action 1. b states that the primary cpolant system temperature should be maintained as low as practical with available equipment. In ITS 3.9.5, a comparable condition exist when SDC train loop requirements are not met. However, ITS 3.9.5 does not contain explicit instructions to maintain the primary coolant system as low as practical with available equipment since this action is beyond the scope of the LCO (i.e., restore compliance with the LCO). Off Normal procedures are used to address alternate ways to maintain the primary coolant system temperature as low as practical when a loss of shutdown cooling exist. As such, CTS Action 1. b has been removed from the CTS and placed in plant procedures. This change is acceptable since these details are not necessary to adequately describe the actual regulatory requirement and placing this information in license controlled documents will not result in a significant impact on safety. This change is consistent with NUREG-1432. LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes associated with this specification .
- Palisades Nuclear Plant Page 3 of 3 01/20/98
- ADMINISTRATIVE CHANGE (A)
ATTACHMENT 3 DISCUSSION OF CHANGES SPECIFICATION 3.9.6, REFUELING CAVITY WATER LEVEL There were no "Administrative" changes associated with this specification. MORE RESTRICTIVE CHANGES (M) M.1 A new specification has been proposed to establish consistency with NUREG-1432. Proposed ITS 3.9.6, "Refueling Cavity Water Level" contains the Limiting Condition for Operation, Applicability, Actions and Surveillance Requirement to ensure a minimum water level is maintained in the refueling cavity during Core Alterations and the movement of irradiated fuel assemblies. The addition of this new specification is appropriate since the minimum refueling cavity water level is an operating restriction that is an initial condition of a design basis event (Fuel Handling Accident Inside Containment) that assumes the failure of a fission product barrier. LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE
- CONTROLLED DOCUMENTS (LA)
There were no "Removal of Detail" changes associated with this specification. LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" . changes associated with this
. specification.
- Palisades Nuclear Plant Page 1of1 01/20/98
- RELOCATED REQUIREMENTS (R)
ATTACHMENT 3 DISCUSSION OF CHANGES SECTION 3.9, REFUELING OPERATIONS R.1 CTS 3.8.ld requires radiation levels in the spent fuel storage area to be monitored continuously during refueling operations. CTS 3.8.2 provides the required actions to be taken when CTS 3. 8. ld is not met. The basis of this requirement is to provide immediate indication when radiation levels exceed a specified setpoint. These radiation monitors do not provide any safety related interlock functions and are not assumed in any Design Basis Event. In addition, these monitors do not meet any criteria in 10 CFR 50.36(c)(2)(ii). Therefore, per 10 CFR 50.36(c)(2)(ii), this specification can be relocated out of the CTS to a licensee controlled document. The requirements associated with the spent fuel storage area radiation monitors are being relocated to the Operating Requirements Manual. Changes to the Opetatirig Requirements Manual will be evaluated using the criteria established in 10 CFR 50.59. R.2 CTS 3.8.lh requires direct communication between personnel in the control room and at the refueling machine whenever changes in*core geometry are taking place. CTS 3. 8. 2 provides the required actions to be taken when CTS 3. 8. lh is not met. Communication requirements allow the control roorri operator to inform the refueling machine operator of any impeding unsafe condition detected from the main control board indicators during fuel movement, as well as, allow for the coordination of activities that require interaction betw~en control foom and contaliiment personnel. The capability to maintain communications between the control room and refueling machine does not meet *any criteria in 10 CFR 50.36(c)(2)(ii). Therefore, per 10 CFR 50.36(c)(2)(ii), this specification can be reloc_ated. out of the CTS to a licensee controlled document. The requirements associated communications .are being relocated to the Operator Requirement Manual: Changes to the Operating Requirements Manual will be evaluated using the criteria established in 10 CFR 50.59. R.3 CTS 3.8.3 prohibits the initiation of refueling operations before the reactor core has decayed for a minimum of 48 hours if the reactor has been operated at power levels in excess of 2 % rated power. The restriction of not moving fuel in the reactor for a period of 48 hours after the power has been removed from the core takes advantage of the decay of the short-life fission products and allows any failed fuel to purge itself of fission gases, thus reducing the consequences of a fuel handling accident. Although this specification satisfies 10 CFR 50.36(c)(2)(ii) criterion 2, the activities necessary prior to commencing movement of irradiated fuel (i.e. reactor head removal, flooding the refueling cavity) ensures that there will at lea.st be 48 hours of subcriticality before movement of any irradiated fuel. Hence, this specification has been relocated as per Industry/NRC agreement during the development of NUREG-1432 .
. Palisades Nuclear Plant Page 1of3 01/20/98
- R.3 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SECTION 3.9, REFUELING OPERATIONS The requirements associated with decay time are being relocated to the Operating Requirement Manual. Changes to the Operating Requirements Manual will be evaluated using the criteria established in 10 CPR 50.59. R.4 CTS 3.8.5 and Table 4.2.2, item 6, require that when spent fuel which has decayed less than one year is placed in the tilt pit storage rack, the bulk water temperature in the tilt pit storage area must be monitored continuously to assure that the water temperature does not exceed 150°F. Monitoring will continue for 24 hours after any addition of fuel to the main pool or the tilt pit, or when a failure of the spent fuel pool cooling system occurs. The bulk water temperature in the tilt pit is higher than the water in the main spent fuel pool because the amount of cooling water flow into the tilt pit is lower for the same amount of fuel assemblies in the main spent fuel pool. For this reason, storage in the tilt pit is limited to irradiated fuel assemblies that have deGayed for at least one year and, assuming the failure of one spent fuel cooling system pump, will not cause the bulk temperature of the tilt pit water to exceed 145°F for the
- nonilal refueling conditions. Although the limit on tilt pit water temperature is intended to prevent damage to spent fuel assemblies, it is not an operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The requirements for the tilt pit do not meet any criteria in 10 CPR 50.36(c)(2)(ii).
Therefore, per 10 CPR 50.36(c)(2)(ii), this specification can be relocated out of the CTS to a licensee controlled document. The requirements associated with the tilt pit are being relocated to the Operating Requirements Manual. Changes to the Operating Requirements Manual will be evaluated using the criteria established in 10 CPR 50.59. R.5 CTS Table 4.2.2, item 5 requires a Refueling System Interlocks functional test prior to refueling operations. The Refueling System Interlocks are designed to prevent damage to fuel assemblies and fuel handing equipment during fuel handling operations. The system interlock test ensures that the equipment used to handle fuel functions as designed and has sufficient load capacity for handling fuel assemblies. Although the CTS does not contain a specific LCO for the Refueling System Interlocks (i.e., no explicit operability requirement exists and no required action for failure to meet the specified test provided), the surveillance requirement for the Refueling System Interlocks was evaluated to the criteria of 10 CPR 50.36 consistent with the evaluation performed for Manipulator Cranes found in NUREG-0212, "Standard Technical i. Specifications for Combustion Engineering Pressurized Water Reactors."
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- R.5 (continued)
ATTACHMENT 3 DISCUSSION OF CHANGES SECTION 3.9, REFUELING OPERATIONS The results of the evaluation concluded that the Refueling System Interlocks did not meet any of the criteria of 10 CFR 50.36 and therefore, could be relocated out of the CTS to a licensee controlled document. As such, the surveillance requirement for Refueling System Interlocks is being relocated to the Operating Requirements Manual. Changes to the Operating Requirements Manual will be evaluated using the criteria established in 10 CPR 50.59 .
- Palisades Nuclear Plant Page 3 of 3 01/20/98
ATTACHMENT 4 PALISADES NUCLEAR PLANT
- SECTION 3.9 - REFUELING OPERATIONS NO SIGNIFICANT HAZARDS CONSIDERATION
- ADMINISTRATIVE CHANGES (A)
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.9, REFUELING O?ERATIONS The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve reformatting, renumbering, and rewording of Technical Specifications. These changes, since they do not involve technical changes to the
- Technical Specifications, are administrative.
This type of change is connected with the movement of requirements within the current requirements, or with the modification of wording which does not affect the technical content of the current Technical Specifications. These changes will also include nontechnical modifications of _requirements to conform to the Writer's Guide or provide consistency with the Improved Standard Technical Specifications in NUREG-1432. Administrative changes are not intended to add, delete, or relocate any technical requirements of the current Technical Specifications. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not
- represent a significant hazards consideration. The following is provided in support of this conclusion.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specification. These modifications involve no technical changes to the existing Technical Specifications. The majority of changes were done in order to be consistent with NUREG-1432. During the development of NUREG-1432, certain wording preferences or English language conventions were adopted. The changes are administrative in nature and do not impact initiators of analyzed events. They also do not impact the assumed mitigation of accidents or transient events. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated .
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- 2.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SECTION 3.9, REFUELING OPERATIONS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The changes will not impose any new or different requirements or eliminate any existing requirements. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in margin of safety?
The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes are administrative in nature and will not involve any technical changes. The changes will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Also, since these changes are administrative in nature, no question of safety is involved. Therefore,
- the changes do not involve a significant reduction in a margin of safety .
MORE RESTRICTIVE CHANGES (M) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve adding more restrictive requirements to the existing Technical Specifications by either making current requirements more stringent or by adding new requirements which currently do not exist. These changes may include additional commitments that decrease allowed outage time, increase frequency of surveillance, impose additional surveillance, increase the scope of a specification to include additional plant equipment, increase the applicability of a specification, or provide additional actions. These changes are generally made to conform with the NUREG-1432. . In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.
- Palisades Nuclear Plant Page 2 of 7 01/20/98
- 1.
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SECTION 3.9, REFUELING OPERATIONS Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. These more stringent requirements do not result in operation that will increase the probability of initiating an analyzed event. If anything the new requirements may decrease the probability or consequences of an analyzed event by incorporating the more restrictive changes. The changes do not alter assumptions relative to mitigation of an accident or transient event. The more restrictive requirements continue to ensure process variables, structures, systems, and components are maintained consistent with the safety analyses and licensing basis. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed changes provide more stringent requirements than previously existed in
- the Technical Specifications. The changes do not alter the plant configuration (no ,
new or different type of equipment will be installed) or make changes in the methods governing normal plant operation. The changes do impose different requirements. However, these changes are consistent with the assumptions in the safety analyses and licensing basis. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in margin of safety?
The proposed changes provide more stringent requirements than previously existed in the Technical Specifications. Adding more restrictive requirements either increases or has no impact on the margin of safety. The changes, by definition, provide additional restrictions to enhance plant safety. The changes maintain requirements within the safety analyses and licensing basis. As such, no question of safety is involved. Therefore, the changes do not involve a significant reduction in a margin of safety .
- Palisades Nuclear Plant Page 3 of 7 01/20/98
- ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.9, REFUELING OPERATIONS LESS RESTRICTIVE CHANGES -REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve moving details (engineering, procedural, etc.) out of the Technical Specifications and into a licensee controlled document. This information may be moved to the ITS Bases, UFSAR, plant procedures or other programs controlled by the licensee. The removal of this information is considered to be less restrictive because it is no longer controlled by the Technical Specification change process. Typically,: the information moved is descriptive in nature and its removal conforms with NUREG-1432 for format and content. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes move details from the Technical Specifications to a licensee controlled document. The changes do not result in any hardware or operating procedure changes. The details being removed from the Technical Specifications are not assumed to be an initiator of any analyzed event. The licensee controlled document containing the removed Technical Specification details will be maintained using the provisions of 10 CFR 50.59 and is subject to the change control process in the Administrative Controls Section of the Technical Specifications. Since any changes to a licensee controlled document will be evaluated per 10 CFR 50. 59, no increase (significant or insignificant) in the probability or consequences of an accident previously evaluated will be allowed without prior NRC approval. Therefore, the changes do not involve a significant increase in the probability or consequences of an accident previously evaluated .
- Palisades Nuclear Plant Page 4 of 7 01/20/98
- 2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.9, REFUELING OPERATIONS Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes move detail from the Technical Specifications to a licensee controlled document. The changes will not alter the plant configuration (no new or different type of equipment will be installed) or make changes in methods governing normal plant operation. The changes will not impose different requirements, and adequate control of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed changes move detail from Technical Specifications to a licensee controlled document. The changes do not reduce the margin of safety since the location of details has no impact on any safety analysis assumptions. In addition, the detail to be transposed from the Technical Specification to a licensee controlled.
- document are the same as the existing Technical Specification. Since any future changes to this licensee controlled document will be evaluated per the requirements of 10 CFR 50.59, no reduction in a margin of safety will be allowed without prior NRC approval.
The existing requirement for NRC review and approval of revisions,. in accordance
. with 10 CFR 50.92, to these requirements proposed for movement, does not have a specific margin of safety upon which to evaluate. However, since the proposed change is consistent with the Combustion Engineering Plants Standard Technical Specification, NUREG-1432 approved by the NRCStaff, revising the Technical Specifications to reflect the approved level of detail ensures no significant .reduction in the margin of safety. Therefore, this change does not involve a significant reduction in the margin of safety .
- Palisades Nuclear Plant Page 5 of 7 01/20/98
- RELOCATED SPECIFICATIONS (R)
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.9, REFUELING OPERATIONS The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve relocating existing Technical Specification Requirements and Surveillances to licensee controlled documents. The Palisades Nuclear Plant has evaluated the current* Technical Specifications using the criteria set forth in 10 CPR 50.36. Specifications identified by this evaluation that did not meet the retention requirements specified in the regulation are not included in the Improved Technical Specifications (ITS) submittal. These specifications have been relocated from the current Technical Specifications to the PSAR or licensee controlled documents referenced in the PSAR. Relocating requirements which do not meet the Technical Specifications criteria to licensee controlled documents allows the Technical Specifications to be reserved only for those conditions or limitations upon reactor operation which are necessary to adequately limit the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety, thereby focusing the scope of the Technical Specifications.
- In accordance with the criteria set forth in 10 CPR 50.92, the Palisades Nuclear Plant has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change relocates requirements and surveillances for structures, systems,
- components or variables which did not meet the criteria for inclusion in Technical Specifications. The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate design basis accident or transient events. The requirements and surveillances for these affected structures, systems, components or variables will be relocated from the Technical Specifications to an appropriate administratively controlled document and maintained pursuant to 10 CPR 50.59. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated .
- Palisades Nuclear Plant Page 6 of 7 01/20/98
- 2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SECTION 3.9, REFUELING OPERATIONS Does the change create the possibility of a new or different kind of accident from Ii.I I any accident previously evaluated? The proposed change does not necessitate a physical alteration* of the plant (no new or different type of equipment will be installed) or change in parameters governing normal plant operation. The proposed change will not impose any different requirements and adequate control of information will be maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The proposed change relocates requirements and surveillances for structures, systems, components, or variables that do not meet the criteria for inclusion in Technical
. Specifications. *The change will not reduce a margin of safety since it has no impact on any safety analysis assumptions. In addition, the relocated requirements and surveillances for the affected structure, system, component, or variable remain the same as the existing Technical Specifications. Since any future changes to these requirements or the surveillance will be evaluated pursuant to 10 CFR 50.59, there will be no reduction in a margin of safety. Therefore, the change does not involve a significant reduction in the margin of safety.
LESS RESTRICTIVE CHANGES (L) The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Changes have been proposed which involve making the requirements in the current Technical Specifications (CTS) less restrictive. A description of the less restrictive change and corresponding No Significant Hazards Consideration are provided on the following pages for each Specification as applicable .
- Palisades Nuclear Plant Page 7 of 7 01/20/98
- LESS RESTRICTIVE CHANGE L.1 NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 SPECIFICATION 3.9.1, BORON CONCENTRATION The Frequency for verifying boron concentration in CTS 3.8. lg has been changed from "on each shift" to *~12 hours." Considering the large volume of water in the primary coolant system (and refueling cavity during Core Alterations), in addition to administrative controls instituted to preclude a boron dilution event, a sampling Frequency of 72 hours is adequatl! to identify slow changes in boron concentration. Furthermore, if a rapid change in boron concentration would occur during refueling operations, an increase in subcritfcal multiplication would be detected by the source range nuclear instrumentation required by proposed LCO 3.9.2, "Nuclear Instrumentation." Therefore, based on the availability of the source range nuclear instrumentation, the large volume of water in the primary coolant system (and refueling cavity during Core Alterations), and administrative control instituted to preclude a boron dilution during refueling operations, performing a verification of the boron concentration on a 72 hour Frequency is considered acceptable. This change is consistent with NUREG-1432
- 1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
- Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change revises the Surveillance Frequency for verifying boron concentration of the primary coolant system and refueling cavity from "on each shift" to "72 hours." Relaxing the Frequency to confirm boron concentration during refueling does not have a detrimental impact on the integrity of any plant structure, system or component. This relaxation will not alter the operation of any plant equipment, or otherwise increase its failure probability. As such, the probability of occurrence for a previously analyzed accident is not significantly increased.
The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The boron concentration assumed upon the initiation of an analyzed event remains unchanged and continues to be periodically verified. The change in sampling periodically does not affect the assumptions of an analyzed event. This change does not affect the performance of any credited equipment since the surveillance is for an assumed parameter. As a result, no analyses assumptions are violated. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event .
- Palisades Nuclear Plant Page 1of2 01/20/98
- 2.
ATTACHMENT 4 NO SIGNIFICANT .HAZARDS CONSIDERATION SPECIFICATION 3.9.1, BORON CONCENTRATION Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those procedures relied upon to mitigate a design basis event. Relaxing the Frequency requirement to confirm the boron concentration does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
- The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Relaxing the Frequency requirement to confirm boron concentration does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety .
- Palisades Nuclear Plant Page 2 of 2 01/20/98
- LESS RESTRICTIVE CHANGE L.1 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION ATTACHMENT 4 CTS 3.17.6 requires two channels of Neutron Flux Monitoring to be Operable below 104 %
Rated Power. CTS 3.17.6.lc requires that shutdown margin be verified within 4.hours, and once each 12 thereafter, whenever one or two Neutron Flux Monitoring channels are inoperable. Proposed ITS 3.9.2 Required Action B.2 only requires a boron concentration verification (which ensures an adequate shutdown margin for existing core conditions) if two source range channels are inoperable, and only requires the verification to be performed once per 12 hours. Both the CTS and ITS require two Operable source range channels.to ensure redundant monitoring capability is available to detect changes in core conditions. With one channel inoperable redundancy will be lost, however, one channel is still available to provide direct means for monitoring core reactivity. Since the capability exists to directly monitor core reactivity conditions with one source range channel inoperable, and core reactivity conditions are stabilized due to the suspension of Core Alterations and positive reactivity additions, there is no need to verify core reactivity conditions by use of chemical analysis other than at the normal sampling frequency of 72 hours. In addition, the accelerated initial
- performance (within 4 hours) of the boron concentration verification when two source range channels are inoperable is excessively restrictive and not warranted. This is based on routine sampling (every 72 hours) and knowledge of stable conditions prior to the loss of the source range channel, and the recognition that a PCS dilution event is detectable through other means such as an uncontrolled increase in the refueling cavity level. This change is consistent with NUREG-1432 as modified by TSTF-96.
- 1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change eliminates the requirement to determine core reactivity conditions by chemical analysis when one source range channel is inoperable, and deletes the requirement to performed an initial boron concentration sample within 4 hours when two source range channels are inoperable. The proposed change does not have a detrimental impact on the integrity of any plant structure, system or component, and will not alter the operation of any plant equipment, or otherwise increase its failure probability. As such, the probability of occurrence for a previously analyzed accident is not significantly increased .
- Palisades Nuclear Plant Page 1of2 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION LESS RESTRICTIVE CHANGE L.1 (continued)
ATTACHMENT 4 The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed changes do not alter the initial conditions for an analyzed event or the availability or functioning of equipment assumed to respond to an analyzed event. Therefore, this change does not* involve a significant increase in the consequences of a previously analyzed event.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or
- those procedures relied upon to mitigate a design basis event. The elimination and relaxation of required actions in the technical specifications does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction. in the margin of safety .
- Palisades Nuclear Plant Page 2 of 2 01/20/98
- LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSJDERATION SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS CTS 3. 8 .1 b requires all automatic containment isolation valves be operable or that at least one valve in each line be closed. Proposed ITS 3.9.d requires each penetration providing direct access from the containment atmosphere to the outside atmosphere be either: 1) closed by a manual valve, automatic isolation valve, blind flange, or equivalent, or; 2) capable of being closed by an Operable Containment Refueling Radiation Monitor. The requirements for containment penetration closure associated with ITS 3.9.3 are relaxed from the requirements of CTS 3. 8 .1 b since they only address components in systems which provide a direct path from the containment atmosphere to the outside atmosphere. The r.eduction in the scope of penetrations from the CTS to the ITS is acceptable since the potential for containment pressurization as a result of an accident is not likely while in MODE 6. Without containment pressurization, only those penetrations which provide a direct path from the containment atmosphere to the outside atmosphere present a potential release path for fission product radioactivity. The less stringent requirements of ITS 3.9.3 continue to ensure that a release of fission product radioactivity within containment as a result of a fuel handling accident will be restricted from escaping to the environment. This change is consistent with NUREG-1432 .
- 1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by failure of plant structures, systems or components. A reduction in the scope of containment penetrations required to be closed or capable of being closed by an Operable automatic isolation signal is not assumed to be an initiator or precursor of any analyzed event. Relaxing the current requirement to only address containment penetrations which provide a direct path from the containment atmosphere to the outside atmosphere does not have a detrimental impact on any plant structures, systems or components since the operation and design of plant structures, systems and components has remained unchanged. As such, the probability of occurrence for a previously analyzed accident is not significantly increased .
- Palisades Nuclear Plant Page 1of6 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS
_LESS RESTRICTIVE CHANGE L.1 (continued) ATTACHMENT 4 The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. Containment _penetrations which do not provide a direct path from the containment atmosphere to the outside atmosphere are not assumed to be credible leakage pathways during the Mode or condition in which this requirement is specified. This is because pressurization of the contaii;iment, which is necessary to providing a driving force for leakage through penetrations in otherwise closed systems, is not assumed in the fuel handling accident analysis which presents the most severe radiological consequences during the condition in which this Specification applies. As such, no analysis assumption are violated. Therefore, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
- The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions.
No change is being proposed to the procedures governing normal plant operation or those procedures relied upon to mitigate a design basis event. Relaxing the requirement relative to containment penetrations which are required to be closed or capable of being closed by an Operable automatic isolation signal does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated .
- Palisades Nuclear Plant Page 2 of 6 01/20/98
- 3.
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS Does this change involve a significant reduction in a margin of safety? ATTACHMENT 4 The proposed change reduces the scope of containment penetrations required to be closed or capable of being closed by an Operable automatic isolation signal to only those penetration which provide a direct path from the containment atmosphere to the outside atmosphere. The margin of safety for the accident postulated during the plant condition in which this requirement applies is determined by the quantity of radioactive material released from the containment atmosphere to the outside atmosphere. The radioactive material is assumed to be released through containment penetrations which provide direct communication between the containment atmosphere and the atmosphere outside containment since leakage through penetrations which contain fluid systems (air and water) is not credible based on the containment pressure remaining at or near standard atmospheric pressure. As such, the proposed change will not result in an increase in the quantity of radioactive material released during an analyzed event. Therefore, this change does not involve a significant reduction in a margin of safety .
. The margin of safety is determined by the design and qualification of the plant
- equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. A reduction in the scope of containment penetrations required to be closed or capable of being closed by an Operable automatic isolation signal does not significantly impact these factors and does not constitute a design change or equipment performance parameter change that would result in a change to the margin of safety. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety .
- Palisades Nuclear Plant Page 3 of 6 01/20/98
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.3, CONTAINMENT PEN"ETRATIONS LESS RESTRICTIVE CHANGE L.2 CTS 3. 8. le requires the containment venting and purge systems be tested and verified Operable "immediately prior to refueling operations." Proposed ITS 3.9.3 and SR 3.0.4 require the Operability of each containment purge and exhaust valves prior to entering the mode of Applicability ( e.g., prior to Core Alterations or the movement of irradiated fuel assemblies within the containment). The CTS has been revised to delete the "immediately prior to refueling operation" requirement. This is acceptable since the ITS provides general rules for the application of surveillance requirements in the technical specifications. SR 3.0.4 establishes the requirement that applicable SRs must be met before entry into a mode or other specified condition in the Applicability. In addition, the specific time frames and conditions necessary for meeting the SRs are specified in the Frequency. Although the phase "immediately prior to refueling operations" implies a conditional type frequency, proposed SR 3.9.3.2 specifies a fixed Frequency of 18 months. The 18 month Frequency maintains consistency with other similar instrumentation (e.g. , ESF AS) and valve testing requirements and is considered acceptable since the containment venting and purge systems also requires the performance of additional SRs to ensure Operability of the isolation function. The SRs for the actuating instrumentation are contained in proposed ITS 3.3.6,
. "Refueling CHR Instrumentation" which also has an Applicability of during Core Alterations and during movement of irradiated fuel assemblies in containment. Therefore, SR 3.9.3.2 is only a system* functional test of the valves to close on a signal while the actuation signal are addressed in ITS 3.3.6. This change is consistent with NUREG-1432.
- 1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. The proposed change revises the time frame for demonstrating the Operability of the containment purge isolation valves prior to entering the Mode of Applicability. Confirming the Operability of the containment purge isolation valves prior to entering the Mode of ApplicabilitY does not have a detrimental impact on the integrity of any plant structure, system or component. This relaxation will not alter the operation of any plant equipment, or otherwise increase its failure probability. The probabiiity that equipment failures resulting in an analyzed event will occur is unrelated to a component which initiates a protective action. As such, the probability of occurrence for a previously analyzed accident is not significantly increased .
- Palisades Nuclear Plant Page 4 of 6 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS LESS RESTRICTIVE CHANGE L.2 (continued)
ATTACHMENT 4 The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints at which these actions are initiated. Performance of the surveillance requirement confirming the operability of the containment purge isolation valves prior to entering the Mode of Applicability ensures that the assumptions of the safety analysis are met and does not affect the performance of any credited equipment. As a result, no analyses assumptions are violated. Based on this evaluation, there is no significant increase in the consequences of a previously analyzed event.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant
- is normally operated or in the setpoints which initiate protective or mitigative actions .
No change is being proposed to the procedures governing normal plant operation or
. those procedures relied upon to mitigate a design basis event. Relaxing the requirement to confirm the Operability of the containment purge isolation valves prior to entering the Mode of Applicability does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated .
- Palisades Nuclear Plant Page 5 of 6 01/20/98
- 3.
NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS ATTACHMENT 4 Does this change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. Relaxing the requirement to confirm operability of the containment purge isolation valves does not significantly impact these factors. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are affected, and no change is being proposed in the plant operational limits as a result of this change. Therefore, this change does not involve a significant reduction in the margin of safety .
- Palisades Nuclear Plant Page 6 of 6 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION- HIGH WATER LEVEL LESS RESTRICTIVE CHANGE L.1 ATTACHMENT 4 CTS 3 .1. 9. 3 allows all flow through the reactor core to be intentionally stopped for up to 1 hour providea, in part, that the core outlet temperature stays ~ 200
- F and two SDC trains are Operable. Proposed ITS 3.9.4 does not contain these additional restrictions. While in MODE 6 with the refueling cavity water level ~647' elevation, an increase in primary coolant system temperature above 200
- F is not an immediate concern. The affects of elevated coolant temperatures at or above the boiling point would eventually challenge the integrity of the fuel cladding, which is a fission product barrier, and lead to a reduction in boron concentration due to boron plating out on components near the area of boiling.
However, due to the relative short time flow is allowed to be suspended (up to 1 hour per 8 hour period), sufficient boiling would not occur such that it would result in a signification reduction in the boron concentration or present a challenge to the fission product barrier. Coolant temperatures above the saturation temperature with no forced circulation become an immediate concern only when the reactor vessel head is installed due to the potential of vapor formations in the primary coolant system loops. The additional restriction in the CTS to maintain two SDC trains Operable when all flow through the reactor core is intentionally stopped is excessively restrictive. since two redundant heat removal methods are still available. That is, when flow is stopped, one SDC train is still required to be Operable and the refueling cavity water level is still required to be ~ 647' elevation thus providing adequate and redundant heat removal capability. This change is consistent with NURE0-1432.
- 1. Does the change involve a significant increase in the probability or consequence of an accident previously evaluated?
Analyzed events are assumed to be initiated *by the failure of plant structures, systems or components. Ensuring the core outlet temperature stays s200°F and that two trains of shutdown cooling (SDC) are Operable when all flow through the reactor core is intentionally stopped, is not assumed to be an initiator or precursor of any analyzed
. event. Ensuring core outlet temperature remains below a specified limit and SDC trains are Operable does not impact the integrity of any plant structure, system or component. As such, deletion of the current requirement will not impact the integrity of any plant structure, system or component. Therefore, the probability of an accident previously evaluated is not significantly increased .
- Palisades Nuclear Plant Page 1of3 01720/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL LESS RESTRICTIVE CHANGE L.1 (continued)
ATTACHMENT 4 The consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event, and the setpoints . at which these actions are initiated. Deletion of the requirement to verify core outlet temperature stays ::;;200"F when flow through the reactor core is temporarily suspended does not alter the assumption of any analyzed event postulated to occur
- while the plant is in MODE 6 and the refueling cavity water level is z647' elevation.
In addition, the availability and functionality of the equipment and systems used in analyzed event during this plant condition have not been altered. Therefore, the proposed change does not involve a significant increase in the consequences of an accident previously evaluated.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a physical alteration of the plant. No new equipment is being introduced, and no installed equipment is being operated in a new or different manner. There is no alteration to the parameters within which the plant is normally operated or in the setpoints which initiate protective or mitigative actions. No change is being proposed to the procedures governing normal plant operation or those procedures reiied upon to mitigate a design basis event. Relaxing the requirement to verify core outlet temperature and SDC train Operability does not have a detrimental impact on the manner in which plant equipment operates or responds to an actuation signal. As such, no new failure modes are being introduced. In addition, the change does not alter assumptions made in the safety analysis and licensing basis. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated .
- Palisades Nuclear Plant Page 2of 3 01/20/98
- SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL 3.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION Does this change involve a significant reduction in a margin of safety? The margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed"limits, and the point at which protective or mitigative actions are initiated. The proposed change eliminates the requirement to maintain core outlet temperature ~200°F and to have two Operable SDC trains during the period when all flow through the reactor core is intentionally stopped. Relaxing this requirement does not impact factors that are related to the margin o,f safety since no changes have been made to plant design, plant equipment or the way in which the plant is operated. Prolong elevated temperatures in the primary coolant system in excess of 212
- F would eventually result in fuel assembly damage.
However, the technical specification continue to limit the duration in which all flow through the reactor core is allowed to be stopped to 1 hour in a 8 hour period. In addition, the technical specifications also require two redundant heat removal method* to be available, they an~; a refueling cavity water level ~647' elevation and one Operable SDC train. As such, the likelihood of fuel damage as a result of elevated temperature is very unlikely. Therefore, the proposed change does not involve a significant reduction in a margin of safety .
- Palisades Nuclear Plant Page 3 of 3 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.5, SDC & COOLANT CIRCULATION- LOW WATER LEVEL LESS RESTRICTIVE CHANGE (L)
ATTACHMENT 4 There were no "Less Restrictive" changes associated with this specification .
- Palisades Nuclear Plant Page 1of1 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION SPECIFICATION 3.9.6, REFUELING CAVITY WATER LEVEL LESS RESTRICTIVE CHANGE (L)
ATTACHMENT 4 There were no "Less Restrictive" changes associated with this specification .
- Palisades Nuclear Plant Page 1of1 01/20/98
ATTACH1\1ENT 5 PALISADES NUCLEAR PLANT
- SECTION 3.9 - REFUELING OPERATIONS MARKUP OF NUREG-1432 TECHNICAL SPECIFICATIONS AND BASES
3.9 REFUELING OPERATIONS
- 3. 9. l LCO 3.9.l Boron Concentration Boron concentrations of the Reast ~~ Coolant System,~
rifweliR~ eaRalQ/and the refueling cavit~shall be maintained.g:hur"thi"' 111111 vspeqr1ea- 11Ytff COBt> (f) G.:t ~e R£F'IA.i:t. ll\J6' Bo 2.0A) C()IJ(. NTR~ nail:) C.T S APPLICABILITY: MODE 6. It!/ 3.n.lo I TC.,... I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration A. l Suspend CORE Immediately not within limit. ALTERATIONS. A.2 Suspend positive Invnediately reactivity additions. AND A.3 Initiate action to Immediately restore boron concentration to within 1 im;t. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY C:..T.f 72 hours SR 3. 9.1.1 3s. t.~ © c.'\J 18 ..... 1 4 . 2..1
NucJe~r Instrumentation 3.9.2
. 3.9 REFUELING OPERATIONS 3.9.2 Nuclear Instrumentation
.; Cha.."""('\ <i..l.S LCO 3.9.2 Two source range (mq;f i tif s <28§4) sha 11 be OPERABLE.
APPLICABILITY: MODE 6. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. l Suspend CORE Irnmediately ALTERATIONS. C:..TS AliD. .
".l .'(. "\.. A.2 Suspend positive Immediately c.rs reactivity additions.
J.n.t... \
- c,rl 3 .~.i.. AND Irnmediately
@(0 B.2 Perform SR 3.9.1.1 7STF-q(o Once per 12 h~r~ -
@tCf f)
~ 3.9-2 Re~95
- fa. thf~p~ ~~t"tf>-, P1(i. (\-t
Nucle~r Instrumentation 3.9.2
- SURVEILLANCE REQUIREMENTS SR 3.9.Z.l SURVEILLANCE Perform CHANNEL CHECK.
FREQUENCY 12 hours SR 3.9.2.2 Perform CHANNEL CALIBRATION. ju~y'months (jJ CEOG STS 3.9-3 Rev 1, 04/07/95
Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the fo 11 owing status: C!5 ------, (;;\ CD a. The equipment hatch closed and held in place by~ujl" 3.'a'. I rDJSc.f<ill* ~ ) bolts; C...~ One door in ~a1r lock closed; and the. c.rn~" '-/ - - - ------ '1' f4'\ d __l. Each penetration providing direct access from the
'2.J containment atmosphere to the outside atmosphere either:
- 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or ,. _
Re.wa/1n~
.2 *. capable of being closed by an OPERABLEAConta~nment (Purge/and Qmaust l}olation Sistem) l*hbl./ Rlld1~.+'1on
/J~rpf APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel asse.mblies within containment .
- C.T 5 ACTIONS CONDITION A. One or R10re A. l REQUIRED ACTION Suspend CORE COMPLETION TIME l11111ediately 3.S2 containment ALTERATIONS.
penetrations not in required status. AHD A.2 Suspend movement of l11111ed 1ate l y irradiated fuel assemblies within containment. 3.9-4 Rev~/07/95
- SECTION 3.9 INSERT
-------------------------------------------------~()'fE:------------------------------------------------------- 'fhe equipment--hatch is only required to be closed when the Fuel Handling Area Ventilation System is not in compliance with LCO 3.7.12, "Fuel Handling Area Ventilation System."
- b. ()ne door in the personnel air lock closed;
--------------------------------------------------~()'fE:--------------------------~---------~------------------
One door in the personnel air lock is orily required to be closed when the equipment hatch is
*closed .
- 3.9-4
Containment Penetrations
. 3.9.3
- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment 7 days penetration is in the required status.
C.TS 3 a.z.. c.. SR 3.9.3.2 ltfiCS ii' a..mirraitt'r. each required.containmen~i~n,ill st valve @ctwateslto tnt '--~~ on an actua 1 or s i mUl ate IS~IU\ Ji~nths C a o signal. '* 7 . Xd.u~J1n Gl'l"ki.l""'"'.lt J.I, ~ i\a.d1~1HI -fJ 1 oP
- - Ntiil ~ _ _ _
\ Q"lf r~.u.1;~cl ~)v fel'\(,trc..+10~.
uni ?c:ia.ttd C!fl"\b1t'\mt.A1° CEOG STS 3.9-5 Rev 1, 04/07 /95
SOC and Coolant Circulation-High Water Level 3.9.4
@.L{ "'ea.<. tot'
- Pri rr'4r '{
.R06~ = fro.,ns 3.9 REFUELING OPERATIONS
- 3.9.4 Shutdown Coaling (SOC) and Coolant Circulation-High Water Level
~ff'c:.i n (Of[i~e:Jc. o.nJ LCO 3.9.4 One SOC ~shall be~in operation. b .
flof- ~ 111
--- ------ ------ -{-ft<;:L".:)._ - - --:NOTE:"- -- -- - - - - - - -- --- - --*- -- -- --
(, The required SDC~may re operation for e.1.S s 1 hour per~J-lioifr' period, provided no aper *ans are
~.1."i.3 permitted that would cause reduction of the R¥cVi r Coolar.t
~' ~~~~~-~~~~ -~~~~~~~~~~ '. ~~ :. ---------------~- '.: '..".:~:~ --------
/
3: ca.1.~ re.~cl.1n 0 .~v1-l-b 01../ 7 ~ e..le.i.u+/on. APPLICABILITY: MOOE 6 with the,..water 1eve1 ~ @ fJ( aboii tho/' top of/ react/f)r ) 6'¥se/ fl;llg~ (continued) A. I .!1\httQ,h. A.t.+10n +o rc.s-tot'"
.soc. ~n -to 0 PlRABll.
ota.+vs <l.l\d 1*~-+16'1._ 3.9-6 Rev 1, 04/07/95
- SECTION 3.9 INSERT The required SBC train may be made inoperable for s: 2 hours per 8 hour period for testing or maintenance, provided one SDC train is in operation providing flow through the reactor core, and core outlet temperature is :S: 200 °F. ,~
- 3.9-6
SOC and Coolant Circulatinn~High Water Level 3.9.4
- ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4 Close a11 containment 4 hours penetrations providing direct access from containment atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY c;r5" Ill- Lf. 2. .2. SR 3.9.4.1 12 hours ill.I'\ 'lf. C-([) CEOG STS 3.9-7 Rev 1, 04/07/95
SOC and Coolant Circulation-Low Water Level 3.9.5 Jtio p ~ fro. 1 ~
- c_;.5 3.1.q~ LCO 3.9.5
- 3. 9 REFUELING OPERATIONS 3.9.5
_ fro.,(is
\~Ji.or = P~ ('('O.r-f Shutdown Cooling (SOC) and Coolant Circulation-Low Water Level Two SOC~ shall be OPERABLE, and one SOC~ shall be
-tro..1'"'
in operation. re..~ud1 f'I~ Cl:w ,;.y APPLICABILITY: MODE 6 with theVwater 1evel < ctfl?t/above )'he to ii of repttor) ltet!sel/fl an§j. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME frtl.\fl ~ ...... A. One SOC ~ A. l Initiate actJ.o.t!!._to Immediately inoperable. restore SOC ~to OPERABLE status. OR A.2 Irrmed iate 1y
- frc..,,..
B. No SOC ~ OPERABLE B. l Suspend operations Immediately or in operation. involving a Prtrno.ry reduction in ([eicID coolant boron concentration. AHO. B.2 Initiate action ..t.L_ Irrmediately restore one SOC Q9:!!e1 frc;.1i-. to OPERABLE status and to operation. (continued) 3.9-8 Rev 1, 04/07/95
SOC and Coolant Circulat;J.1-Low Water Level 3.9.5
- ACTIONS CONDITION REQUIRED ACTION
'lr.\t!Q.~ ClC.~101'\ +o COMPLETION TIME B. (continued} B.3 ,t1ose all containment penetrations providing direct access from containment atmosphere to outside atmosphere.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY C.iS 3.IA.3 SR 3.9.5.l 12 hours c C.TS T~ y.z..z. 1'11{ SR 3. 9. 5. 2 Verify corre~t breaker aHgnment _and 7.days ClJ TB(.. '-(.2.. "l.
*indicated power available to the required SOC pump that is not in operation.
"It IL(
Je..n+"f O"l"lt ~CC. f~,n \S In; 6~,..°ti~fl
- a. od c.~rc:.~~trnu fr1 rr;o.ry &c/4f'\t- ~+ Q.,
flow ~~
- at ** ? it>>o *'a'.f ('('...
CEOG STS 3.9-9 Rev 1, 04/07/95
~;+(
Refueling~Water Level 3.9.6
- ACTIONS 3.9 REFUELING OPERATIONS
(.p.v;f.'( 3.9.6 RefuelingvWater Level
,Jc,uJ LCO 3.9.6 lfu.
c;.~ 'fi Cf\'t.; I fJO e,.\J APPLICABILITY: ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
~v'rfY A. Refueling~water level A. l Suspend CORE Immediately not within limit. ALTERATIONS.
M!l2
- -rsTf*7-o A.2 8W2 Suspend movement of irradiated fuel assemblies within
*containment.
Immediately A.3 Initiate ac on to restore ref eling ed1ately cavity wa r level to within l i it. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 3.9-10 Rev 1, 04/07/95
~*H Ref~elingAWater Level
.. 3.9.6
- ACTIONS SR 3.9.6.1 C.a.111+y Verify refuelingvwater level is~ 2 ove e op o reac or vesse 24 hours CEOG STS 3.9-11 Rev 1, 04/07/95
Boron Concentration B 3.9.l
- B 3.9 REFUELING-OPERATIONS B 3.9.1 Boron Concentration BASES f .* Pr1'rnQ.r-y .
BACK_GROUND The limi t/ on the boron concentrations of theaffiaefeft Coolant System ~S), (t!jj?retyelrnaztanre:i and refueling cavity the re~1.1.~l1~~ durin refuelin ensur s that t e reactor remains c\)erA.:t'o 1u su cr1t1ca ur1ng MODE 6. boron concentration' is the soluble boron concentration in the coolant in each of these volumes having direct access to the reactor core during refueling. The soluble boron concentration*offsets the core reactivity and is measured by chemical analysis of a represent~tive'!ff;.~ de'~\\u\ ~ A~~ sample of the coolant in each of the volumes. The r *n Spec.\ ~\l.O..t; cM I.I)
"~e~l~i~ lol\S .
cf'P.S"(]oron concentratiori)limit is rocedures ensure the s ecified boron concentration in order PIA .. t to ma1n ain an overa core reactivit of k <
- during
/iS:"i fuel handl in~, with control e n a sem "'and .rods ,\.J..J ~ **~ fuel assembl1es assumed to be 1n t e most adverse I~!.~}--, c.onfiguration (least negative reactivity) al lowed by~ p~"t @-
- L __ .____pr~cedures. ~ . .
~(. Pc.!ir~ . 11 uc.~ fkcr-."T (GDC l6 of 10 QfR so/ AopeiJ61x D requires that two
- e.si~I"\. c..rdc:.,...a...
('1f:::- independent reactivity control systems of different design principles be provided (Ref. 1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. ine Chemical and Volume Control System
\...'.:::) (CVCS)~is ~system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration.
p The reactor is brought\to shutdown conditions before beginning operations tct open the reactor* vessel for refueling. After the(!CS is cooled and de~ressurizedrailcl\ the vessel head is unbolte<tt~the head 1s Slewly remove~ (fojlj tl>e teij(e i 161 caV/tY>. The (refllel jlig cifial. ifid & 1.s... refuel inq cavit~ then flooded with borated water from s.a~ in~cn _ _--.t-.he..,.,.refueling water tank into the open reactor vessel by
. :=Jvit~feeding or by the use of the ~U[<fown Coplin# csQtD
~'5tem~uml?,!:) S'Pc.o+ fu:L.Coi.1i\q or ~1::.-/.(t"~'"",.;f*"'p "s 1 ~~'1 1~.iec.tio" U
- Pu111415, Ot ~""1 The pumping action of the SOC System in the (Hts and the fuw./J'.
natural circulation due to thermal driving head<<) in the reactor vesse 1 (anf the tefye I Wj cavity )mix the added {con~ed)
\cEpG/SZSl B 3.9-1 Rev 1, 04/07/95
- Po.J 1Sc;.4c Noc.kc..r ~"t
- SECTION 3.9 INSERT During evolutions where plant procedures allow manipulation of control rods or where conditions could result in inadvertent control rod withdrawal, such as reacto~ vessel head removal, the boron concentration must be sufficient to assure that kerr will remain~ 0.95 without taking credit for the negative reactivity provided by the control rods (ie., assuming aw rods fully withdrawn). During evolutions where the control rods are inserted, plant procedures do not allow manipulation of control rods, and conditions do not exist that could result in inadvertent rod withdrawal, such as MODE 6 operations with the Upper Guide Structure in place (other than during head removal). Therefore, credit may be taken for the negative reactivity provided by the control rods when determining the boron concentration necessary to assure that kerr will remain ~ 0.95 .
- B 3.9-1
Boron Concentration B 3.9.1
- BASES BACKGROUND (D.vrfj l ~ntrated boric acid with the water in the refueling (continued) ~~. The SOC System is in operation during refueling (see LCO 3.9.4, "Shutdown Cdoling (SOC) and Coolant Circulation-High Water Level, and LCO 3.9.5, "Shutdown 8
Cooling (SOC) and Coolant Circulati n-Low Water Level") to rovid forced circulation in the an t*ass1s ln p maintaining t oron c a lo in the (Et
~~ ~9 cij(ar~ and the refueling cavityt_t~~~~~;:.~~~~
~ AU..
t:AP~ Pla"-T The required boron concentration and the!lii\lit\refueling procedures that demonstrate the correct tueTloading plan (including full core mapping) ensure the keff of the core will remain s 0.95 during the refueling operation. Hence, at least a 5% AJr../k margin of safety is established during
- refueling
- During refueling, th; water volume in~he spent fuel pool, the transfer canal, (thi refUil int can~:'> the refueling cavity, and the reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these volumes.
The limiting boron dilution accident analyzed-occurs in MOOE 5 (Ref. 2). A detailed discuss~* of~hj~event is provided in B l.l:t, "SHUTDOWN MARGI f*vgfi 2 I -
"?*"
~ .AcS)iorpn concentration satisfies Criterion 2 of @"""t~...,e (@j ~""Jta t'gmiiif;> - I D C.. f ((. 5 0 . j ~ ( () ( 2.) 0 f
LCO The LCO requiresfthat a minimum boron concentration be maintained in th~(~SJ"thi rifueijrjg cap(r~ and refueling (9) cavity while in MOOE 6. The boron concentration limit specified [Pf W cot'.B) ensures a core keff of s O. 95 is 0 CEOG STS B 3.9-2 Rev 1, 04/07/95
Boron Concentration B 3.9.l
- BASES LCO (continued) M maintained during fuel handling operations. Violation of could lead t_o an inadvertent critical_ity.@ufi!)ID APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical. The required boron concentration ensures a ~~f ~ 0.95. Above MODE 6; lsrf-1:,U> LCO 3.1.1 "SHUTDOWN MARGIN Suri ~
c 3. . 2 " UT WN RG . 2bva* , " ensure that an adequate amount o negative reactivity is available to shut down the reactor and to maintain it subcritical. ACTIONS 1\.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent ~on maintain int th~~ in comp l i ance with the LCO. If t e boron concen rahon of any coolant volume f... in the.:~ t)i't rftuel {nglcanil ~or the refueling cavity is less than its im1t, a 1 operat1ons involving CORE ALTERATIONS or positive reactivity additions must be
- suspended irrmediately .
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving.a component to a safe positionJ,of {)~/ (.oo/d<<.an ~f: h1e. ~iJlan-+ IJofurn<. 9.,r +-""(. f(Jrp()~c. o-t p~~ fe-Mfc.t'oofll'r- un trol... y In addition to irrmediately_, suspending CORE--AbTERATIONS or positive reactivity additions-,- boration to restore the concentration must be. initiated irrmediately. In determining the required combination of boration flow rate and concentration, there is no unique design basis event that must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for~conditions. fla.t"I+) CEOG STS B 3.9-3 Rev 1, 04/07/95
Boron Concentration B 3.9.l
- BASES ACTIONS A.3 (continued)
Once boration is initiated, it must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration. SURVEILLANCE SR 3. 9.1.1 REQUIREMENTS A minimum Frequency of once every 72 hours is therefore a reasonable amount of time to verify the boron concentration of representative samples. The Frequency is based o~ operating experience, which has shown 72 hours to be adequate. FSA-~ Sl!:c.1->~ s.1
$i.
REFERENCES 1. 6Ci CFR AiiDendil A, }iDC 26;1
- 2. FSAR, Section Jel. 1"1. 3
~
CEOG STS B 3.9-4 Re.v 1, 04/07/95
Nuclear Instrumentation B 3.9.2
- B 3.9 REFUELING OPERATIONS B 3.9.2 Nuclear Instrumentation BASES
/
t: C.fo.~ncl< ( Nr.
- a 1I o3 ._, J N[ -~2/0'I) . .
©
- APPLICABLE SAFETY ANALYSES
~Sourc., ra."~< c.h~"~ ... 1.s Two OPERABLE~are required to provide a signal to alert the operator to unexpected changes in core reactivity such as by a boron dilution accident or an improperly loaded fuel assembly. The safety analysis of the uncontrolled boron dilution accident is described in Reference 2. The analysis of the uncontrolled boron dilution accident shows that normally available SHUTDOWN MARGIN would be reduced, but there is sufficient time for the operator to take corrective actions. .
. I -' lc.5
- IV.x.b, * . tftle gei(S"lsatis@ Criterion 3 of We Ni(C P@icv /Stafemerit).
1toJttruWl1.f\in+lf,(\ .{ ID C..f R SD.3to(c. Xz.)
~50LJret. f c.ncx<- c..hc.."' "c..ls LCO This LCO requires two~ OPEAABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity.
(continued) B 3.9-5 Rev l, 04/07/95 lo bt.c:fAIL, e"'chc.hci,r.I m.istPro\i1cic 11fS1Jcl indicn;t1cn a..f'lcl a.tica4t one. ot the. tr.Uo c.h:inflGb mvtt frt'ILhdt an a.L.Jd1&lc Cwnt r~:k. .P()ric.+11'>1'1 It\ tr.t Cbti"trol. r'Oolt\. lfic.rc.torc., With no O.t.Jd ib/.c. . C,i)"",. rq,i<, tvflo1-i&f\ frOM a.+ Jeo.at Ol'\c. Cfunnt.l, b~h r:iotrc.e. nl.t\~ U-0."'l'K.ls Wovlrl
~ 11'\ o pc.nJ>k.,
Nuclear Instrumentation B 3.9.2
- BASES APPLICABILITY 4 swrc.t. ra."~' C.1-o"f\"'\.S In MODE 6, the~ must be OPERABLE to deter'!111 n@) changes in core reactivity. There is no other direct means available to check core reactivity levels.
dc.+e.c.t In MODES 2, 3, 4, and st the installed source range Cha.l"\1"\-c.l.s (deticto121 ~itt 5ff'51MJ! are required to be. OPERABLE by LCO 3.3*.~ '-]t$11ist;l'.Yminoht1orx Snutamm ; c:r. 1~ M~r.J ru.1.,_
~
IY\&fl1tor*"ti Ch°"""'"
- LS
.If'\ moDl I, ein( §JIU. !tc."'G' d:;a::nn ..l i.s ~1J1f"td '1 /.L() 3.3.i, ~Ma.+t Shilt-~~ 11 St~*
ACTIONS A.I and A.2 With only one<Si!}OPERABLE, redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions, CORE ALTERATIONS and positive reactivity additions must be suspended i1T111ediately. Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position~ or (10f'f1\Ptl.. CAoldOUI" ~+ 'th., Golllo.f\t lloiuf'l'I'(. fo .. "tri< p1,;r-p4p.. of ~.t~.,._
. ft.. ""tJe-rof~ ~ n1t"<jL.. *
.Ll.
.____,,-cs C."-ci.nnc.L.
With no~ OPERABLE, action to restore aClriQfii1f>i!> to OPERABLE status shall be initiated i1T111ediately~nce initiated, action shall be continued until~~~is
- restored to OPERABLE status.
OPERABLE, there is no direct means of detecting changes 1n core reactivity. However, since CORE ALTERATIONS o~< and ositive reactivit additions are not to be ma~de the core reactivity con it1on is stabilized until th are OPERABLE. This stabilized condition is determined y performing SR 3.9.1.1 to verify that the required boron concentration exists. o..nd rt\vJ.*"' eo,,/1t1
~Ct. Pc.r 12- ll The Complet* Time of~hours
- su ficient to obtain and Pr1rr-tJ..rf analyze a* r c or. coolant am le for boron concentration.ta.nd l\JlllUJl:.l:.!il:!IYJW.lil"1...lW:!i!.nwc~J!SJULl:lli.llU.I ensures that unplanned changes in boron concentration would be identified. The 12 hour Frequency is reasonable, considering the low probability of a change in core reactivity during this period.
(continued) CEOG STS B 3.9-6 Rev 1, 04/07/95
Nuclear Instrumentation B 3.9.2
- BASES (continued)
SURVEILLANCE REQUIREMENTS SR 3.9.2.1
. SR 3.9.2.1 is the performance of a CHANNEL CHECK, which.is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core condition~Changes in fuel loading and core geometry can result rn significant differences,,
between source range channels, but e~ch channel shouTd be consistent with its local conditions. SR 3.9.2.2 REFERENCES 1. Q~l: sojAppendif A: GC }3* ~DC" 26yGDC 2a:l.§)
- 2. FSAR, Section ).,k_l!.f,3 CEOG STS B 3.9-7 Rev 1, 04/07/95
Containment Penetrations B 3.9,3 B 3.9 REFUELiNG OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGP.OUNO During CORE ALTERATIONS or movement of fuel assemblies within containment with irradiated fuel in containment, a
/1.r. N"lbCL.. s., rib Q.u1dc.C\k release of fission product radioactivity within the
~;c 0..JJ[JLJ Med wh I d.. WJ i_L containment will be restricted from escaping to the envi ro.nment when the LCO requirements are met. In MODES 1,
- fu)<; t+ ,"o.. rt.-1~-t. at
- 2, 3, and 4, this is accomplished by maintaining containment
. ('o.dt~ac..+-r.ic Mo.. h.r1c!.P ft OPERABLE as described in LCO 3.6.l, "Containment." In
~.ODE 6, the potential for containment pressurization as a fr-. "l. ~ r'I t---..... l "" .i)\C.. rct o. t mo.r Pk re.
- result of an accident is not likely; therefore, requirements
/h(.l"C. ~rt..) no S"~v\((..fl\c,(lfS a.re..
to isolate the containment from .the outside atmosphere can be less stringent. i'he LCO requirements are referred to as
/J t 1Pul.J;. hJ d11 vfi~,,,
~
,¢,,1 d
/] 'wl'\1:,.fr-GA r/*N.
"containment closure" rather than "containment OPERABILITY."
Containment closure means that all potential escape paths (0 -ri1kr~d, ar!t-closed. or capable of being closed. Since there is no
~-'p'-ot~ential for containment pressurization, the Appendix J 1 eakage criteria and tests are not required '1t: 10 t.fR so !_. _ _ _ _ ©____ _J The containment. serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100.
Additionally, the containment structure provides radiat1on shielding from the fission products that may be present in the containment atmosphere following accident conditions. The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containme 1 a must be held in place y at east our o s. Good eng1neering practice dictates tha~ the bolts required by this LCO be approximately equally spaced. The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during HODES 1, 2, 3, and 4 operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is req~ired. During periods of shutdown when containment (continued) aEJij$!S) B. 3.9-a Rev 1, 04/07/95
- ~1.:: ,JcJ~~lorrl
- SECTION 3.9 IN SERI During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment with the equipment hatch removed, the OPERABILITY requirements of the Fuel Handling Area Ventilation System must be met. These OPERABILITY requirements are provided in LCO 3.7.12, "Fuel Handling Area Ventilation System" .
- B 3.9-8
Containment Penetrations B 3.9.3
- BASES BACKGROUND (continued}
closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open
- for extended periods when frequent containment entry is.
necessary. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always
!;r.f\,)s!l...;_j} ____3~.!-~n .~l_?sed.~ . .
f'2.\ The requirements on containment penetration closure ensure
'\:::::!.) that a release of fission product radioactivity within containment will be restricted from escaping to the environment. The closure restrictions are sufficient to restrict fission product radioactivity release from containment due to a fuel* handling accident during refueling.
(IJUS"f.LT 3 !. and all The m* ipurge system is no used in HOOE 6. All four [8]
- ch valves are secur in the closed position.
~Jther containment penetrations that provide direct access from containment atmosphere to outside atmosphere
;t>-a.+ a.r71 ntit C.O..fo.I:,(, o.f, be1n() <:.loScJ t,t °'" O P~&ll J\~+xuj ~n-b.lnnitrl 1-h1f Ka.c.hitHn (.)' 1"cV (continued)
CEOG STS B 3.9-9 Rev 1, 04/07/95
- SECTION 3.9 INSERT 1 An exception, *however, is provided for the personnel air lock. It is acceptable to have both doors of the pe~sonnel air lock opened simultaneously provided the equipment hatch is opened.
INSERT 2 and venting the containment is accomplished using the Clean Waste Receiving Tank (CWRT) vent line. INSERT 3 During CORE ALTERATIO NS or movement of irradiated fuel assemblies within containment with either the Containment Purge and Vent System in operation, or the CWRT aligned for containment venting, the associated isolation valves must be capable of being closed by an OPERABLE channel of radiation instrumentation required by LCO 3.3.6, "Refueling Containment High Radiation Instrumentation. "
- B 3.9-9
Containment Penetrations B 3.9.3
- ItJ5fZ.7 J.
APPLICABLE SAFETY ANALYSES (D CAv1ty fL~.,,~.,I bf
.(1h~Oilt\,-4-1~0 I../ g - - - I
&_~llUnc.C'\tf ffiz.nJal) -~~""""-~,........--....."""""" ... I 2~ +H~~ ~~~~~~~~~~~~,..:,r.,;~:-rz..:Tr~~-:Ti~~~~-:-::':-::-'
I LCO to {5) Re.fvcli,.,"'1, Lc)0~1l'l'M*"~I
. ~j (f (~~10..fthJ.
J () SfrtJ MC..r.. ki.flon, 1 1rJ:Ste13 (continued) CEOG STS 8 3.9-10 Rev 1, 04/07/95
- SECTION 3.9 INSERT 1 Containment penetrations "that provide direct access from containment atmosphere to outside atmosphere" are those which would allow passage of air containing radioactive particulates to migrate from inside the containment to the atmosphere outside the containment even though no measurable differential pressure existed. Specifically, they do not include penetrations which are filtered, or penetrations whose piping is filled with liquid.
INSERT 2 Containment penetration isolation is not required by the fuel handling accident to maintain offsite doses within the guidelines of 10 CPR 100, but operating experience indicates that containment isolation provides significant reduction of the resulting offsite doses. Therefore, the Containment Penetrations satisfy the requirements of Criterion 4. of 10 CPR 50.36(c)(2) .
- INSERT 3 do not assume a specific closure time for the valves in these penetrations since the accident analysis makes no specific assumptions about containment closure time after a fuel handling accident.
INSERT 4-- LCO 3.9.3.a is modified by a Note which allows the equipment hatch to be opened if the Fuel Handling Area Ventilation System is in compliance with LCO 3.7.12. LCO 3.9.3'.b is modified by a Note which allows both doors of the personnel air lock to be simultaneously opened provided the equipment hatch is opened. With both doors in the personnel air lock opened and the equipment hatch opened, the Fuel Handling Area Ventilation System maintains the atmosphere in the spent fuel pool area at a negative pressure relative to the auxiliary building (adjacent to the personnel air lock) and containment building. In the event of a fuel handling accident inside containment, any radioactivity released to the containment atmosphere will either remain in the containment or be filtered through the Fuel Handling Area Ventilation System. As such, with the equipment hatch removed, and both personnel air lock doors opened, the consequences of a fuel handling accident in containment would not exceed those B 3.9-10
Containment Penetrations B 3.9.3 BASES LCO (continued) Al'eLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when therP is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1, "Containment." In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status. ACTIONS A. I and A.2 SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed iosition i~ in t~~~ition. The Surveillance on the op n purge ad) (@ valve will demonstrate that the valves are not oc e rom closing. Also, the Surveillance will demonstrate that each valve operator has motive power, which will ensure each valve is capable of being closed by an 1n u,, 1r.oiA.+c.J fbnc.~+io4 w"1ch Qrou1"'-c. a.. d1lc.o\ Po.¥n ~rG\'Y'\. fN, C61'1-\G\n~ a.+moi~V\. +6 +'°" 6vtlidt C\+Mo8P~V"Ci (continued) CEOG STS B 3.9-11 Rev 1, 04/07/95
Containment Penetrations B 3.9.3
- BASES SURVEILLANCE REQUIREMENTS R(Jl.Jw"*?\ Coniu. m~l"lr 1
SR 3.9.3.1 (continued)
~~~~~~Jwtomi'tic C.CSntainPiint Pir9e a&J exha!Jtt isolaflom 1-fi ~ l<lt'ct -t-1 on j.::.. _ _ _ __.
O* T~e Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment. The Surveillance interval is selected to be corrmensurate with the normal duration of time to tomplete
@) fuel handlin o erations. © (continued)
CEOG STS B 3.9-12 Rev 1, 04/07/95
Containment Penetrations B 3.9.3
- BASES (continued}
REFERENCES I. on SE-000200 -001, Rev. , 0 t. l, July 1!¥117 (:D CEOG STS B 3.9-13 Rev 1, 04/07/95
SOC and Coolant Circulation-High Water Level (/]\ o a B 3.9.4
'-.:!.) JYAC.1cr : r ri IV\l:U' y
- . - P.c.s ~ fc.s B 3.9 REFUELING OPERATIONS ~ 00 ~ *- 1rcLi" B 3.9.4 Shutdown Cooling (SOC) and Coolant Circulation-High Water Level
. l.~-- .. 1 BACKGROUND The purposes of the SOC ystem in.l.MODE 6 are to remove decay heat and sensible he f the <R"ia¢00?.> Cool ant System p t'.BCS), as required b 4 to provide mixing of borated "taolant, to provide suf 1 1ent coolant circulation to minimize the effects of a boron dilution accident, and to Pr1m~ry revent boron stratification Ref. 1). Heat is removed from p--.!h.!r S by circu at1ng etc ~coolant through the SOC heat exchanger(s), where the eat is transferred to the Component Cooling Water System via the SOC heat exchanger(s). The coo an 1s en re urne o . S via ~Scold leg(s).
Operation of the SOC System for normal cooldown or decay heat r~moval is manually accomplish~d from the control room. Th eat removal rate is adjusted by controlling the flow of cto *coolant through the SOC heat exchanger(s) and - bypassing the heat exchanger(s). Mixing of the~ fri~ c~ is maintained by this continuous circulation of I
~coolant through the SOC System *
(continued) B 3.9-14 Rev 1, 04/07/95
SOC and Coolant Circulation-High Water Level B 3.9.4
- BASES APPLICABLE SAFETY ANALYSES (continued)
SOC and C~ant&rcuhMon Hilh ~ter Level satisfies Criterion(t)of LJ NRC _l icy _at_enO.
'I ~10 CFR 50.3tp Cc)(.z).
-fra1.;,
LtO is required for decay heat removal Only one SOC is required i.C..1 n because the volume o water above the reactor vessel flange. provides backup decay heat removal capability. At least one SOC ~must b~ in operation to provide: tf'C1.ll'\ 1..ofERft~ a."J
- a. Removal of decay heat;
- b. Mixing of borated coolant to minimize the possibility of a criticality; and
- c. Indication of reactor coolant temperature.
fz:'\ o JfJ' o~
'11..Y An OPERABLE SOC p an* SOC pump, a heat exchanger, valves, piping, 1 ruments, and controls~~~ an J ft~
OPERABLE fl ow path and to detenni ne the CfowJM)'iemperature. The fl ow path starts in. one of th.e 'S hot legs and is returned t°'i <f.Jre(l[S cold 1egl'. ..1. b , .1..
---p -fc.>o N6"\'~ . (ICT e. '"' o ero...1 ,~
The LCO is modified by
- Notef~ allows the required
~i,, _Qperat i ng SOC,crD6iiJto re e m c for up to 1liour in eachShour period, provided no oper!~ons are p permitted that would cause a reduction of the~S boron
- concentration. Boron concentration-*reduction 1s prohibited because uniform concentration distribution cannot be ensured without forced circulation. This pennits operations such as a core mapping or alterations in the vicinity of the reactor r vessel not 1eg nozzles, an'Cf~s to soc isolation valve testing *. During this 1 hour period, decay heat is removed c 1rc.ul.J;..ti 6YJ by naturat,.@ev§C}1 P'D to the large mass of water in the
@ liN~-------re_fu_e_11_*n_g_ca_v_1t_y-J.1' . . & ~~on 47 (continued)
CEOG STS B 3~9-15 Rev l, 04/07/95
- SECTION 3.9 INSERT Note 2 allows the required SDC train to be made inoperable for~ 2 hours per 8 hour period for testing and maintenance provided one SDC train is in operation providing flow through th.::
reactor core, and the core outlet temperature is ~ 200°F. The purpose of this Note is to allow the heat flow path from the SDC heat exchanger to be temporarily interrupted for maintenance or testing on the Component Cooling Water or Service Water Systems. During this 2 hour period, the core outlet temperature must be maintained ~ 200°F. Requiring one SDC train to be in operation continues to ensures adequate mixing of the borated coolant .
- B 3.9-15
SOC and Coolant Circulation~High Water Level B 3. 9.. 4
- BASES APPLICABILITY (continued)
L:O 3.9.5, "Shutdown Cooling (SOC) Low Water Level." ACTIONS ~'" SOC([!Qg)requirements are met by having one SOC ~f"<<.1~ o OPERABLE and in operation, except as permitted in e Note to the LCO.
~il fro.,~ . .
If SOC~requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can occur through the addition of water with a lower boron f concentration than that contained in the~S. Therefore, actions that reduce boron concentration shall be suspended i11111ediately . (continued) CEOG STS B 3.9-16 Rev 1, 04/07/95
- SECTION 3.9 INSERT A. l If one required SDC train is inoperable or not in operation, actions shall be immediately initiated and continued until the SDC train is restored to OPERABLE status and to operation. An immediate Completion Time is necessary for an operator to initiate corrective actions .
- B 3.9-16
SOC and Coolant Circulation-High Water Level B 3.9.4
- BASES ACTIONS (continued)
+o The. f!nlJlt?riftllt, SURVEILLANCE REQUIREMENTS
(]) REFERENCES 1. FSAR, Section5_LQJ.. &. I a."J /L./.J
- CEOG STS B 3.9-17 Rev 1, 04/07/95
SOC and Coolant Circulation-Low Water Level f'Lj\. B 3.9.5
\J Kto.c.~~ . . ~iltrol'"\/
P.c.s : Pc..s B 3.9 REFUELING OPERATIONS ~oof : t('b.*" B 3.9.5 Shutdown Cooling {SOC) and Coolant Circulation--Low \rlater Level BASES BACKGROUND (.,. /
. Prirna-,y---,
- APPLICABLE SAFETY ANALYSES If thelfeictop"coolant~temperature is not maintained below 200° F, boiling of the rfnt;Jor> coolant could result. This could lead to inadequate cooling of the reactor fuel due to the resulting loss of coolant in the reactor vessel.
Additionall boilin ffj@ttof)coolant could lead to a re uct1on in boron concentr tion in the coolant due to the boron plating out on components near the areas of the boiling activity, and because of the possible addition of water to the reactor vessel with a lower boron concentration than is required to keep the* reactor subcritical. The loss o 6'{iic1~ coolant and the reduction of boron concentration 1n th ~!!ct§e coolant would eventually challenge the 1n egrity of the fuel cladding, which is a fission product barrier. Two trains of the SOC System are requi*red to be
@* OPERABLE, and one train is required to be in operation in MODE 6, with the a e re.+vw~ &.t.J1~y ( \L..::.~:=.:..~:c.:::..::..:...~=laM' to prevent t is cha 11 enge.
C.UO.hr- .fc:'J.c.L <: t'h<...
~ ~1 ft e.lw:J..°" SOC and Coolant Cjrcylation-Low Water Level satisfies Criterion of (the NRC Policy Statement;)
L{ ~ ( 0 C.f ({ 5 a..H~ ( C.) (LI .
~
(continued) B 3.9-18 Rev 1, 04/07/95
SOC and Coolant Circulation~Low Water Level B 3.9.5
- BASES (continued)
- a. Removal of decay heat;
- b. Mixing of borated coolant to minimize the possibility of a criticality; and ACTIONS th~ lJLfl ~t O(' ~r~~
(continued) CEOG STS B 3.9-19 Rev 1, 04/07/95
SOC and Coolant Circulation....:..Low Water Level B 3 .9 .. 5
- BASES ACTIONS
{continued)
.fL.1 tfb.\.r . .+rc..,;..s If no SOC ~is in operation or no SOC~ are OPERABLE, there will'-ife"ilo forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron CQncentrations can occur by the addition of water with lower boron concentration than that contained in the<,'.BX:""'S:'°' f Therefore, actions that reduce boron concentration shall. be suspended i11111ediately.
u +ro*~.., +,. ... ,....J. If no SOC <!Qfi)is
- eration or no SOC~ are OPERABLE, action shall be initiated is8&ij~ and continued without interruption to restore one ~to OPERABLE status and operation. Sjnce tha.Af@is in Conditions A and B f-te.'.,.,
concurrently, the restoration of two OPERABLE SOC~ and one operating SOC~ should be accomplished expeditiously.
+-a.,*,,
B.3 S~c.. trc..,~ 1mm*d1Clfc.lf *
© If nofRRQ?"loomis in operation, all containme-*)penetrations providing direct access from the containmenifatmos here to
- @
- scc.-t~
1 *"
~he the "de' atmosphere must be closed w1 o requirements not met, the pate t e coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded~
a ou s With exists for he
'"¢ '.
SURVEILLANCE REQUIREMENTS (continued) CEOG STS B 3.9-20 Rev 1, 04/07/95
SOC and Coolant Circulation~Low Water Level B 3.9.5
- BASES SURVEILLANCE REQUIREMENTS SR 3.9.5.1 (continued) trv.1" In addition, during operation of the SOCG:22E:)with the water level in the vicinity of the reactor vessel nozzles, the SOC
-fr'o,;.. ~flow rate detennination must also consider the SOC pump suction re uirements. The Frequency of 12 hours is su 1c1ent, cons1 er1ng the flow, temperature, pump contro*1, and alarm indications available to the operator to monitor the SOC System in the control room.
Ver'fication that the require loops are OPERABLE a op ration ensures that loops an be placed in oper as n eded, to maintain decay h t and retain forced irculation. The Frequenc~ of 12 hours is consid red reasonable, since other a inistrative controls re available and have proven to be acceptable by o erating experience. SR 3.9.5.2 ~* y
-~
*verification that the required pump is OPERABLE ensures that an additional SOC pump can be placed in operation, if needed, to maintain decay heat removal andtr&\ttnr>coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump .
The Frequency of 7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. (}) REFERENCES I. FSAR, Sectio~ JV 4. (,,.I a.~ I 1(.3, CEOG STS B 3.9-21 Rev 1, 04/07/95
r'FiV*W Refueling~Water Level B 3.9.6
- B 3.9 REFUELING OPERATIONS
~,+.y B 3.9.6 RefuelingfWater Level BASES
© BACKGROUND APPLICABLE During core alterations and during movement of irradiated SAFETY ANALYSES fuel assemblies, the water level in the(riltlelii§ caniJl"inil) refueling cavity-is an initial condition design parameter in the analysis of the fuel handling accident in containment r[;-. k
~w *C.
~
tS C..ffro~ 1 - postulated by Regulatory Guide 1.25 (Ref. 1). A minimum
*water level of 23 ft (Regulatory Position C.1.c of Ref. 1)
-----=>allows a decontamination factor of 100 (Regulatory
+o a,,... tlt.i.,t).-\--1of\ Position C.1.g of Ref. 1) to be used in the accident o~ 0~1 ~t ~~ ~~i~h: ~o~~~i~~d1n!h~!1~:!:~e~r~~ ~~= ~!~~=r~~n ci~~~ing 1 0 gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain~ of the total fuel rod iodine 131 inventory (Ref. J) . 11.. '10 2.
_ The fuel handling accident analysis inside containment is _ described in Reference 2. With a minimum water level ~ '-{~ and a minimum decay time of ~'1ours prior to fuel
---.....:~~~~h':":an=-=d;.f'_;ling, the analysis and)test opHtam$) demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and o~ite doses are maintained withi!J,(al liiWabW [~ fl!f
- ty.
ter level satisfies Criterion 2 of~
~i-E.~f-:a~em~e~nI:'-: t o cf f... 5 o.3" Cc. t' 2.)..
(continued) C OG/STS B 3.9-22 Rev 1, 04/07/95
- Pa1iJ~rJ "'"°" n~~
CAc.;w RefuelingAWater Level B 3.9.6
- XU)J) +rnA +~
BASES LCO (continued) f e>v1+/
~m.inimum refueling~water
(~ ff\... &
\I 1..n tt ~ l-c.va.+101°'
level {(if"(irt:J=-:f~tl....,.a....,.b-o-vfl""""7"""t...,...h_e_r_e_a_Ct,,_o_i:J cVeisel f,!an§i)is req4ired to ensure that the radiological consequences of a ostulated fuel handling accident inside (f1dJ....!nt.. o-P 10 c.FR.. tW-} con alnm n ar 1t rn ac a e 1m1t"S'-as v1 e
©APPLICABILITY CORE ALTERATION~--
r1 a c ng an un ate i of con ol rod A'rive iiafts, an w en mov1 ng ue ass em 11 esfn eprelence o I 11 C.ori+a.1n ~""t irradiated fuel assemblie~. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel is not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3. 7. , "Fuel @fd§j> Pool Water Level. n 14 Spo:l ACTIONS A.I and A.2 .[.be.low +rt-<. &;4i .Pt Uc.vo.+tonJ With a water level of < ~ft abov the top of he re to (Vissel {I ange";) a11 opera 1ans 1nvo v1 ng or movemen of irradiated fuel assemblies shall be suspended i11111ediately to ensure that a fuel handling accident cannot occur. The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position.
- ion to i11111ediately s pending CORE ALTE mov nt of irradiated fuel action to restore r, fueling cav* y water level must b initiated i11111ediate *
(continued) CEOG STS Rev 1, 04/07/95
f.A.,.\.'1 Refueling~Water Level B 3.9.6
- BASES (continued)
SURVEILLANCE REQUIREMENTS SR 3.9.6.1 ~(4.JP~Yd*"o ~ t~ (J-17 ~.f e,/c.tA+tO"\ Verification of a minimum. water level of r1i2ft aboyi{ the tOl6) (of tj(e reattor vesse1 flangEJ ensures that the design basis
- for the postulated fuel handling accident analysis ~
rf::fueling operations is met. Water at the required~ e/e.l):l.tlOr'\ (above;tne top/of the rjl~tof Vessel ffingi>limits the* t<./ consequent.es of damagedue rods that are postulated to result from a fuel handling accident inside containment (Ref. 2). . The Frequency of 24 hours is based on engineering judgment and is considered adequate in view of the large volume of water and the nonnal procedural controls of valve positions, which make significant unplanned level changes unlikely. REFERENCES 1. Regulatory Guide 1.25, March 23, 1972. (0 2. FSAR, Section ~. IL/. ft CD (3. /NUREGJ)soo. sedion 15h.4J GI 10 @R 100. ~.) CEOG STS B 3.9-24 Rev 1, 04/07/95
ATTACHMENT 6 PALISADES NUCLEAR PLANT
- SECTION 3.9 - REFUELING OPERATIONS JUSTIFICATION FOR DEVIATIONS FROM NUREG-1432
- Chana:e DiscussiOn ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.1, BORON CONCENTRATION Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications' were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
- 1. The brackets have been removed and the proper. plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatiCal preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.
- 3. The requirement/statement has been deleted since it is not applicable to this
- 4.
facility. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specification.
- 6. The Palisades Nuclear Plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to the "Palisades Nuclear Plant design criteria." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs .
- Palisades Nuclear Plant Page 1of2 01/20/98
- Discussion ATTACHMENT 6 JUSTIFICATION FOR :Q:EVIATIONS SPECIFICATION 3.9.1, BORON CONCENTRATION
- 7. TSTF-136 combines ISTS 3.1.1 and ISTS 3.1.2 and renumbers the remaining specifications in Section 3.1. In addition, TSTF-136 modifies the Bases of Specification 3. 9 .1 to correct references to Section 3 .1. this change is consistent with NUREG-1432 as modified by TSTF-136.
- 8. A statement has been added to clarify that a normal cooldown of the coolant
- volume for the purpose of system temperature control does not constitute a positive reactivity addition.
- 9. Additional information has been included in the Bases Background discussion to clarify the limit on boron concentration relative to the reactivity conditions established by plant procedures and to avoid a potential misunderstanding of the LCO requirement. The Bases states that "Plant procedures ensure the specified boron concentration in order to maintain an overall core reactivity of keff < 0.95 during fuel handling, with control rods and fuel assemblies assumed to be in the most adverse configuration (least negative reactivity)
- allowed by plant procedures." The revised wording explicitly states that the refueling boron concentration must be sufficient to assure keff remains < 0.95,.
without taking credit for control rods, during plant evolutions such as reactor vessel head removal. Credit may be taken for the negative reactivity provided by the control rods (when determining the necessary boron concentration) only* during those evolutions allowed by plant procedures, that do not allow control* rod manipulation or a condition to exist which could result in an inadvertent rod withdrawal, such as Mode 6 operations with the Upper Guide Structure in place .
- Palisades Nuclear Plant Page 2 of 2 01/20/98
- Change SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications' were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes o~
intent.
- 3. The requirement/statement has been deleted since it is not applicable to this
- 4.
facility. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specification.
- 6. The initial performance of SR 3. 9 .1.1 within 4 hours of entry into Condition B
- has been deleted. The accelerated performance of this SR is not warranted based on routine performances of this SR (every 72 hours), and knowledge of stable conditions prior to the loss of the source range channel. Secondarily, PCS dilution events are recognizable through other means such as uncontrolled increases in pool water level. This change is consistent with NUREG-1432 as modified by TSTF-96.
- 7. The Note associated with SR 3.9.2.2 which states that "Neutron detectors are excluded from Channel Calibration" has been deleted. Containing this information in a Note is not necessary since the CTS definition and proposed ITS definition for Channel Calibration state that neutron detectors are excluded from the Channel Calibrations .
- Palisades Nuclear Plant Page 1of2 01/20/98
ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.2, NUCLEAR INSTRUMENTATION Discussion
- 8. The LCO section of the Bases for ITS 3. 9. 2 is revised to add descriptive information regarding the OPERABILITY requirements of the source range neutron flux channels consistent with the assumptions used in the boron dilution accident analysis. This change is considered an editorial enhancement*
only and does not involve a technical issue.
- 9. A statement has been added to clarify that a normal cooldown of the coolant volume for the purpose of system temperature control does not constitute a positive reactivity addition .
- Palisades Nuclear Plant Page 2 of 2 01/20/98
- Change SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS Discussion ATTAC1'ENT 6 JUSTIFICATION FOR DEVIATIONS Note: 1'_!1.is attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic justifications. are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.
- 3. The requirement/statement has been deleted since it is not applicable to this
- 4.
facility. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specification. The allowance to have the equipment hatch removed and to have both personnel air lock doors open simultaneously during CORE ALTERATIONS or the movement of irradiated fuel assemblies in containment has previously been approved by the staff and is part of the CTS. As a condition of this allowance, the fuel handling area ventilation system is required to be in service. The requirement is annotated by a Note in LCO 3.9.3 and is specifically addressed by proposed ITS LCO 3. 7 .12, Fuel Handling Area Ventilation System.
- 6. In the Background section the word "filtered" has been included as a method for ensuring containment closure. The addition of this word reflects current
- plant design which allows the equipment hatch between the containment building and fuel handling area to remain open during refueling activity .
- Palisades Nuclear Plant Page 1of4 01/20/98
- Change SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
- 7. ~-discussion of the methods used to purge and vent the containment has been provided to reflect plant design. In MODEs 1-4, the containment is vented using the vent line from the Clean Waste Receiving Tank (CWRT). The CWRT vent line provide a direct path from the containment atmosphere to the outside atmosphere and contains two isolation valves in series which isolate the CWRTs in the event of high containment pressure or high radiation levels in containment. The Containment Purge and Vent System is comprised, in part, of a 12 inch purge supply line and two 8 inch vent lines which remain isolated in MODEs 1-4. In MODEs 5 or 6, the containment is typically purged and vented using the Containment Purge and Vent System. Although not typically used, the CWRT vent line may also be used as a containment vent path. Therefore,
. the Bases of ITS 3. 9*. 3 has been revised to account for the methods used for containment purge and vent since these penetrations provide a direct path from the containment atmosphere to the outside atmosphere.
- 8. The Background section of the Bases for ITS 3. 9. 3 is revised to clarify the specific provisions used in approving equivalent isolation methods. This change eliminates the potential for misinterpretation of the requirements for approving these methods and provides consistency in the application and understanding of these requirements. The inappropriate reference to a GPU Safety Evaluation is also omitted. This is an editorial enhancement.
- 9. The Applicable Safety Analysis section in the Bases has been revised to delete the discussion that the fuel handling accident analyzed the dropping of a handling tool or a heavy object onto other irradiated fuel assemblies. The context of this statement, as used in NUREG-1432, reflects the events that were considered in the Standard Review Plan (NUREG-0800). The fuel handling
. accidents analyzed in UFSAR Section 14.19 only evaluated the drop of a spent fuel assembly or the dropping of a fuel assembly on another fuel assembly.
- 10. The Applicable Safety Analysis section in the Bases has been revised to delete the discussion which defines the phrase "well within" as it pertains to the .
guidelines of 10 CFR 100 and defined in the Standard Review Plan. The acceptance limits for offsite radiation exposure resulting from a fuel handling accident have previously been approved by the NRC staff and are part of the current licensing basis .
- Palisades Nuclear Plant Page 2 of 4 01/20/98
ATTAC1'ENT 6
. JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS Change Discussion
- 11. In the Bases, the LCO and SR 3.9.3.2 discussions have been revised to eliminate reference to valve closure times. In the fuel handling accident analysis, no credit is taken for the closure time of the valves in the containment _
purge and vent penetrations. In addition, the reference to actuation response ' time in SR 3.9.3.2 has been omitted since this test is not part of the current licensing basis.
- 12. The Bases for SR 3.9.3.2 is revised to omit the reference to manual initiation since it is not required by the SR. This change eliminates the potential for misinterpretation of the SR requirements and provides consistency in the application and understanding of these requirements. This is an editorial enhancement only to provide consistency between the TS and the Bases, and does not involve a technical issue.
- 13. ITS SR 3.9.3.2 and the Bases Background discussion have been modified to more clearly reflect the requirements of the LCO. ITS LCO 3.9.3.d.2 applies to all containment penetrations which provide direct access from the containment atmosphere to the outside atmosphere that are capable of being closed on a high radiation signal. The wording of SR 3.9.3.2 has been revised to match the wording in the LCO in order to eliminate the potential for a misapplication of the surveillance requirement by only testing the containment purge and vent valves. In addition, SR 3.9.3..2 has been modified by a Note which clarifies that only unisolated containment penetrations are required to be tested since penetration which are already isolated are in their accident position and satisfy the requirement of LCO 3.9.3.d.l. Verification of penetrations which are closed by an automatic isolation valve is performed in accordance with SR 3.9.3.1.
- 14. The Bases Background section has been revised to address MODE 5. The ISTS version discusses the importance of "Containment Operability" in MODES 1, 2, 3, and 4 and the need for "containment closure" in MODE 6, but does not mention why containment operability or containment closure are not required in MODE 5. The proposed change clarifies that there are no specific requirements
*for containment penetrations in MODE 5 which would result in a release of radioactive material to the containment atmosphere .
- Palisades Nuclear Plant Page 3of4 01/20/98
- Change SPECIFICATION 3.9.3, CONTAINMENT PENETRATIONS Discussion ATTAC1'ENT 6 JUSTIFICATION FOR DEVIATIONS
- 15. The sentence in the Bases for SR 3.9.3.1 which states "A surveillance before the start of refueling operation will provide two or three surveillance verifications during the applicable period for this LCO" has been deleted. This statement does not provide an explanation or justification for the performance o(
this surveillance, or for its associated frequency. Stating that the performance of the SR will provide two or three surveillance verifications during the applicable period of the LCO was found to be misleading since it is not the intent to perform this SR two or three times, but only to ensure it is performed prior to Core Alterations or movement of irradiated fuel within containment, and every 7 days. In general, fuel movements and Core Alterations are accomplished in less than 7 days .
- Palisades Nuclear Plant Page 4 of 4 01/20/98
- SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL ChanKe Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on ~e "NUREG MARKUPS." The first five justifications were used generically throughout the markup of the NUREG. Not all generic
-justifications are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of
- intent.
- 3. The requirement/statement has been deleted since it is not applicable to this
- 4.
facility. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specification.
- 6. ITS 3. 9 .4 has been revised to require the operating SDC loop (train) to be Operable in addition to being in operation. This change was made to establish consistency with the Bases of ISTS 3.9.4 which describes an Operable SDC loop and with NUREGs-1430 and 1431 which also require one SDC loop to be Operable and in operation. This change is consistent with NUREG-1432 as modified by TSTF-146 .
- Palisades Nuclear Plant - Page 1 of 3 01720/98
-* SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION- HIGH WATER LEVEL Change Discussion ATTACHMENT 6 JUSTIFICATION FOR QEVIATIONS
- 7. The Applicability of ITS 3.9.4 has been revised from "MODE 6, with the water level ~23 ft above the top of reactor vessel flange" to "MODE 6, with the refueling cavity water level ~647 ft elevation." For the Palisades plant, a:J.
elevation of 647 ft is approximate to a level 23 ft above the reactor vessel flange as described in NUREG-1432. This change has previously been found acceptable in the NRC Safety Evaluation related to CTS Amendment No. 173 dated October 10, 1996 which states, "The staff has verified the Palisades upper guide structure design and confirmed that a refueling cavity water level of 647 feet equates to a water level 23 feet above the top of the reactor vessel flange." Corresponding changes have also been made to the Required Actions, SRs and Bases as applicable.
- 8. The Palisades Nuclear Plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CFR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to the "Palisades Nuclear Plant design." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CFR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs.
- 9. In the Applicable Safety Analyses, the criterion of 10 CFR 50.36 satisfied by Specification 3.9.4, "SDC and Coola~t Circulation - High Water Level" has been changed from Criterion 2 to Criterion 4. A loss of shutdown cooling in MODE 6 is not an accident or transient evaluated in the safety analysis for the Palisades plant. Inclusion of this specification in the ITS is based on operating experience which has shown the ability to maintain an Operable and operating SDC train to be significant to public health, and safety and is consistent with the Commission's Safety Goal and Severe Accident Policies. This change is also consistent with NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants" and NUREG-1431, "Standard Technical Specifications, Westinghouse Plants. "
- Palisades Nuclear Plant Page 2 of 3 01/20/98
** *1 ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.4, SDC & COOLANT CIRCULATION - HIGH WATER LEVEL Discussion
- 10. The basis for the Completion Time of Required Action A.4 has been revised to be more succinct. The 4 hour Completion Time was based, in part, on the low probability of the coolant boiling in this time. The current wording of "without incurring the additional action of violating the containment atmosphere" is somewhat nebulous. The revised wording is consistent with the terminology used in other ISTS NUREGS.
- 11. In ITS 3.9.4, an additional exception to the LCO (Note 2) has been provided to allow both the operating and operable SDC loop to be inoperable for the purpose of performing necessary testing and maintenance on valves or piping which are common to both SDC trains. The LCO exception maintains the SDC flow path while permitting testing and maintenance on shared systems.
This exception is necessary since the plant was licensed with a common set of heat exchangers for the Emergency Core Cooling System (SDC is a subsystem) and Containment Spray System. This exception is consistent with the current licensing basis and has previously been found acceptable in the
- 12.
staffs Safety Evaluation related to Amendment No. 161 to facility operating license No. DRP-20 dated August 12, 1994. LCO 3.9.4 Note 1 contains an exception to the LCO but has wording which is inconsistent with the wording of the LCO. This Note and its respective Bases discussion have been revised to provide consistent wording with the requirement being excepted. This change is consistent with NUREG-1432 as modified by TSTF-153 .
- Palisades Nuclear Plant Page 3 of 3 01/20/98
- SPECIFICATION 3.9.5, SDC & COOLANT cmCULATION - LOW WATER LEVEL Chan1:e Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justifications' were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.
- 3. The requirement/statement has been deleted since it is not applicable to this
- 4.
facility. The following requirements have been renumbered, where applicable, . to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specification.
- 6. The Applicability of ITS 3.9.5 has been revised from "MODE 6, with the water level <23 ft above the top of reactor vessel flange" to "MODE 6, with the refueling cavity water level < 647 ft elevation." For the Palisades plant, an elevation of 647 ft is approximate to a level 23 ft above the reactor vessel flange as described in NUREG-1432. This change has previously been found
- acceptable in the NRC Safety Evaluation related to CTS Amendment No. 173 dated October 10, 1996.which states, "The staff has verified the Palisade.s upper guide structure design and confirmed that a refueling cavity water level of 647 feet equates to a water level 23 feet above the top of the reactor vessel flange." Corresponding changes have also been made to the Required Actions, SRs and Bases as applicable .
- Palisades Nuclear Plant Page 1of3 01/20/98
- SPECIFICATION 3.9.5, SDC & COOLANT CIRCULATION - LOW WATER LEVEL Chana:e Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS
- 7. SR 3. 9. 5 .1 which requires a verification that the required SDC loop is Operable and one SDC loop is in operation has been revised .. Proposed ITS SR 3. 9. 5 .1 requires a verification that one SDC train is in operation and circulating primary coolant at a flow rate equal to or greater than 1000 gpni.
The proposed wording of SR 3. 9. 5 .1 is identical to the proposed wording in SR 3.9.4.1 since both SRs are intended to verify primary coolant is circulating through the core at a specified flow rate. Confirmation of a specified flow rate ensure train Operability since an Operable SDC train consist of an SDC pump, heat exchanger, valves and Operable flow path. Verification that the second SDC train is operable is required by SR 3.9.5.2.
- 8. The Palisades Nuclear Plant was designed prior to issuance of the General Design Criteria (GDC) in 10 CPR 50. Therefore, reference to the GDCs is omitted and appropriately replaced by reference to the "Palisades Nuclear Plant design." The Palisades Nuclear Plant design was compared to the GDCs as they appeared in 10 CPR 50 Appendix A on July 7, 1971. It was this updated discussion, including the identified exemptions, which formed the original plant Licensing Basis for future compliance with the GDCs.
- 9. In the Applicable Safety Analyses, the criterion of 10 CPR 50.36 satisfied by Specification 3.9.5, "SDC and Coolant Circulation -Low Water Level" has been changed from Criterion 2 to Criterion 4. A loss of shutdown cooling in MODE 6 is not an accident or transient evaluated in the safety analysis for the Palisades plant. inclusion of this specification in the ITS is based on operating experience which has shown the ability to maintain two Operable SDC trains and one SDC train in operation to be significant to public health and safety, and is consistent with the Commission's Safety Goal and Severe Accident Policies. This change is also consistent with NUREG-1430, "Standard Technical Specifications, Babcock and Wilcox Plants" and NUREG-1431, "Standard Technical Specifications, Westinghouse Plants. "
- Palisades Nuclear Plant Page 2 of 3
\
01/20/98
- SPECIFICATION 3.9.5, SDC & COOLANT CIRCULATION - LOW WATER LEVEL Change Discussion ATTACHMENT 6 JUSTIFICATION FOR D~VIATIONS
- 10. The LCO Bases discussion for ITS 3.9.5 has been revised to clarify that an SDC train is still Operable even when both SDC pumps are aligned to the Safety Injection Refueling Water Tank for filling and draining the refueling cavity, or for required testing. During these operational conditions, cooling is still provided to the reactor core by the operating SDC pump, and as such, is an acceptable plant configuration. This change is consistent with NUREG-1432 as modified by TSTF-21. Subsequent to the approval of TSTF-21, the Bases of ISTS SR 3.9.5.1 was revised by TSTF-21 R.1 to delete the phase "and circulating reactor coolant. " The intent of this change was to accommodate filling and draining evolutions of the reactor cavity using the SDC pumps while continuing to meet the requirement of the SR to circulate reactor coolant. The change proposed by TSTF-21 R.1 was :p.ot adopted in the ITS since circulation of the priniary coolant (as discussed in the LCO portion of the Bases) is assumed to occur during these evolutions.
- 11. The Completion Time for Required Action B.3 has been changed from "4 hours" in the ISTS to "Immediately" in the ITS. Required Action B. 3 requires all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere be closed when there are no SDC loops Operable or in operation. The ISTS Bases states that 4 hours is reasonable (for closing penetrations) based on the low probability of the coolant boiling in that time. ISTS Required Action B.3 is a new action for the Palisades Plant. Current procedural requirements direct the operators to determine the time until the PCS temperature reaches 200°F based on decay time, initial PCS temperature and refueling cavity water level, and to establish containment closure before that temperature is reached. As such, the 4 hours specified in the ISTS may not be conservative for some plant conditions. Therefore, proposed ITS 3.9.5 Required Action B.3 has been change to "Immediately" initiate containment closure on a loss of shutdown cooling .
- Palisades Nuclear Plant Page 3 of 3 01/20/98
- Chan Ke Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.6, REFUELING WATER LEVEL Note: This attachment provides a brief discussion of the deviations from NUREG-1432 that were made to support the development of the Palisades Nuclear Plant ITS. The Change Numbers correspond to the respective deviation shown on the "NUREG MARKUPS." The first five justification8 were used generically throughout the markup of the NUREG. Not all generic justifications are used in each specification.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Deviations have been made for clarity, grammatical preference, or to establish consistency within the Improved Technical Specifications. These deviations are editorial in nature and do not involve technical changes or changes of intent.
- 3. The requirement/statement has been deleted since it is not applicable to this
- 4.
facility. The following requirements have been renumbered, where applicable, to reflect this deletion. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specification.
- 6. The LCO of ITS 3. 9. 6 has been revised to delete reference to a water level
";;::23 ft above the top of reactor vessel flange" and replaced with "~647 ft elevation." For the Palisades plant, an elevation of 647 ft is approximate to a level 23 ft above the reactor vessel flange as described in NUREG-1432. This change has previously been found acceptable in the NRC Safety Evaluation related to CTS Amendment No. 173 dated October 10, 1996 which states, "The staff has verified the Palisades upper guide structure design and
- confirmed that a refueling cavity water level of 647 feet equates to a water level 23 feet above the top of the reactor vessel flange." Corresponding changes have also been made to the Required Actions, SRs and Bases as applicable .
- Palisades Nuclear Plant Page 1of2 01/20/98
- Discussion ATTACHMENT 6 JUSTIFICATION FOR DEVIATIONS SPECIFICATION 3.9.6, REFUELING WATER LEVEL
- 7. The Background, Applicable Safety Analysis, and LCO sections of the Bases have been revised to clarify the offsite dose release from a fuel handling accident in containment as it pertains to the guidelines in 10 CFR 100. The
- acceptance limits for offsite radiation exposure resulting from a fuel handling
- accident have previously been approved by the NRC staff and are part of the current licensing qasis.
- 8. Required Action A.3 and its associated Bases discussion have been deleted.
Required Action A.3 stipulates an action which is to be taken after the plant has left the MODE of Applicability (i.e., suspension of Core Alterations and movement of irradiated fuel assemblies within containment). In addition, restoring compliance with the LCO is always an option and is not generally stated in the ISTS unless it is the only Required Action. This change is consistent with NUREG-1432 as modified by TSTF-20 . Palisades Nuclear* Plant Page 2 of 2 01720/98
I . - --, -
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' . I MPROVED 1
I l I ltECHNICAL . : $PECIFICATIONS ~ I I I I I I AL-I SA-C>-ES UC LEAR LANT i
- . ~* i,
~
- ___________ ----~ volume 19 CHAPTER 4.0 '~Enen.w
PALISADES NUCLEAR PLANT CONSUMERS ENERGY Docket 50-255 Conversion to Improved Technical Specifications License DPR-20 INTRODUCTION: CHAPTER 4.0. DESIGN FEATURES A. ARRANGEMENT AND CONTENT OF THIS CHAPTER OF THE CHANGE REQUEST This Chapter of the Technical Specification Change Request (TSCR) proposes changes to those Palisades Technical Specifications addressing -DESIGN FEATURES. These changes are intended to result in requirements which are appropriate for the Palisades Nuclear Plant, but closely emulate those of the Standard Technical Specifications, Combustion Engineering Plants, NUREG 1432, .Revision 1, Chapter 4.0. This discussion and its supporting information frequently refer to three sets of Technical Specifications, and to two groups of discussions associated with the proposed changes; the following abbreviations are used for clarity and brevity: CTS - The Palisades Current Technical Specifications, ITS - The Palisades Improved Technical Specifications, ISTS - NUREG 1432, Revision 1. DOC - Discussions of Change; these discussions explain and justify the differences between the requirements of CTS and ITS.
- JFD - *Justifications for Deviation; these discussions explain the differences.
between the requirements of the ITS and the ISTS. Six attachments are provided to assist the reviewer:
- 1. Proposed ITS Chapter 4.0 pages
- 2. This Attachment is not applicable to Chapter 4.0
- 3. A set of alJ those CTS pages which contain requirements associated with those in ITS Chapter 4.0, marked up to show the changes from CTS to ITS, and arranged by specification in the order in which the requirements occur in ITS. This attachment also includes a DOC for each change.
Each change from CTS to ITS is classified in the following categories: ADMINISTRATIVE - A change which is editorial in nature, which only involves movement of requirements within the TS without affecting their technical content, or clarifies CTS requirements. MORE RESTRICTIVE - A change which only adds new requirements, or which revised an .existing requirement resulting in additional operational restrictions.* RELOCATED - A change which* only moves requirements, not meeting the 10 CFR 50.36(c)(2)(ii) criteria, from the CTS to the Operating Requirements Manual (which has been included in the FSAR by reference). 1
DESIGN FEATURES INTRODUCTION: CHAPTER 4.0. A. ARRANGEMENT AND CONTENT OF THIS CHAPTER OF THE CHANGE REQUEST (continued) LESS RESTRICTIVE - REMOVAL OF DETAIL (LA) - A change in which certain details from otherwise retained Specifications are removed from the ITS and placed in the Bases, FSAR, or other licensee controlled documents. LESS RESTRICTIVE - A change which deletes any existing requirement, or which revises any existing requirement resulting in reduced operational restriction.
- 4. No Significant Hazards Analyses for the changes from CTS to ITS.
An individual No Significant Hazards Analysis is provided for each Less Restrictive change; generic No Significant Hazards Analyses are provided for each of the other categories of change.
- 5. ISTS Chapter 4.0 marked to show the differences between ISTS and ITS.
- 6. JFDs for the differences between ISTS and ITS.
B. REFERENCE DOCUMENTS This Chapter of the TSCR is based on the following reference documents:
- 1. CTS as revised through Amendment 178.
- 2. The following TSCRs which are currently under review by the NRC:
- a. Containment System, submitted on March 26, 1997.
- 3. ISTS, as revised by Industry Generic Changes (TSTF) approved as of October 15, 1997.
- 4. The following changes to ISTS which are currently under review by the NRC:
- a. None .
- 2
INTRODUCTION: CHAPTER 4.0. DESIGN FEATURES
- c. THE UNIQUE PALISADES NUCLEAR PLANT FEATURES AFFECTING THIS CHAPTER Palisades has several physical, analytical, and administrative features which
*differ from those newer CE plants upon which the ISTS were based. Palisades was
*the first CE plant designed and built. Its design and licensing preceded the issuance of_ the General Design Criteria so that, in some aspects, its physical.
systems are not like those of newer plants; its Technical Specifications preceded the issuance of Standard Technical Specifications (STS) so that *Leos, Actions, and Surveillance Requirements are not coordinated as they would be for a STS plant. Palisades has purchased all its core reloads from Siemens Power Corporation (or its predecessors), therefore, reload analyses and the associated core physics parameters, as well as certain Safety Analyses are not like those plants using all CE fuel and analyses as were modeled in the ISTS. 11 11 D. THE DIFFERENCES BETWEEN CTS 0PERATING CONDITIONS" AND ITS MODES 11 The CTS definitions of plant operating conditions have been replaced with the oper~tion Mode definitions used in ISTS. In several instances the name for a CTS defined "operating condition" is the same as that for an ISTS "Mode," but the definition differs. CTS contain the following definitions for operating conditions:
- 1.
2. The POWER OPERATION condition shall be when the reactor is critical and the neut~on flux power range instrumentation indicates greater than 2% of RATED POWER. . The .HOT STANDBY condition shall be when Tave is greater than 525°F and any of the CONTROL RODS are withdrawn and the neutron flux power range instrumentation indicates less than 2% of RATED POWER.
- 3. The HOT SHUTDOWN condition shall be when the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Specifi ca ti on 3.10 and Tave is greater than 525°F.
4.* The COLD SHUTDOWN condition shall be when the primary coolant is at SHUTDOWN BORON CONCENTRATION and Tave is less than 210°F. * .
- 5. The REFUELING SHUTDOWN condition shall be when the primary coolant is at REFUELING BORON CONCENTRATION and Tave is less than 210°F .
- 3
INTRODUCTION: CHAPTER 4.0. DESIGN FEATURES D. THE DIFFERENCES BETWEEN CTS "OPERATING CONDITIONS". AND ITS "MODES" (continued) ITS contain the following definition table for Modes:
% RATED AVERAGE PRIMARY REACTIVITY THERMAL COOLANT MODE TITLE CONDITION . POWER(a) TEMPERATURE (k,ff) (oF) 1 Power Operation ~ 0.99 > 5 NA 2 Startup ~ 0.99 ,;: 5 NA 3 Hot Standby < 0.99 NA ~ 300 4 Hot Shutdown l*l < 0.99 NA 300 > T.,. > 200 5 Cold Shutdown<*> < 0.99 NA ,;: 200 6 Refueling 1'l NA NA NA
- (a) *
(b) (c) Excluding decay heat. All reactor vessel head closure bolts fully tensioned. One or*more reactor vessel head closure bolts less than fully tensioned. E. . MODE CHANGES USING CTS "OPERATING CONDITIONS" VERSUS ITS "MODES"
- 1. CTS "Power Operation" is essentially equivalent to ITS "MODE 1." Each represents a condition with the reactor critical and the turbine generator in operation. The only effective difference is the power level which separates that Condition or Mode from the next lower one~ During plant startup, the plant must meet all CTS "Power Operation" or ITS "MODE 1" LCOs before the turbine generator is placed on the line; similarly, during plant shutdown, the-plant-exits CTS "Power Operation" or ITS "MODE 1" when the turbine generator is no longer in service. Therefore, this change in definition will have no operational effect.
- 2. CTS "Hot Standby" is similar to ITS "MODE 2." Each represents a condition with the reactor critical, or nearly so, and the turbine generator shut down.
During plant startup, the plant must meet all CTS "Hot Standby" or ITS "MODE 2" LCOs before a reactor startup is started; during plant shutdown, the plant exits CTS "Hot Standby" or ITS "MODE 2" when the reactor is shutdown. CTS action statements requiring that the plant be placed in Hot Standby" are 11 effectively equivalent to ITS Actions requiring the plant be placed in "MODE 2." Therefore, this change in definition will have no operational
- effect.
4
INTRODUCTION: CHAPTER 4.0. DESIGN FEATURES E. MODE CHANGES USING CTS "OPERATING CONDITIONS" VERSUS ITS "MODES" (continued)
- 3. CTS "Hot Shutdown" and ITS "MODE 3" are similar at their upper temperature boundary. During plant shutdown, the plant exits CTS "Hot Standby" or ITS "MODE 2" when the reactor is shutdown. CTS action statements requiring that the pl_9nt be placed in 11 Hot Shutdown 11 are effectively equivalent to ITS Actions requiring the plant be placed in "MODE 3. 11 CTS "Hot Shutdown 11 and ITS 11 MODE 3" are quite different at their lower temperature boundary; CTS "Hot Shutdown" is exited when Tave drops below 525°F, ITS "MODE 311 is not exited until Tave drops below 300°F.
- 4. CTS does not provide a defined term for the condition when Tave is between 525°F and 210°F (the upper bound for CTS "Cold Shutdown 11 ) .
- 5. CTS "Cold Shutdown" is essentially equivalent to ITS "MODE 5." Each represents a condition with Tave below boiling. There is no technical significance to the difference between the CTS 210°F and the ITS 200°F.
CTS Section statements requiring that the plant be placed in Cold Shutdown"
*are effectively equivalent to ITS Actions requiring the plant b~ placed in "MODE 5. 11 Therefore, this change in definition will have no operational effect.
- 6. CTS 11 Refueling Shutdowri" is essentially equivalent to ITS "MODE 6." Each,
- F.
when taken ~ith other definitions and LCO requirements, represents a condition with the reactor at least 5% shutdown. Therefore, this change in definition will have no operational effect. THE MAJOR CHANGES FROM CTS (as modified by pending TSCRs) TO ITS
- 1. The discussion of Containment Design Features in CTS Section 5.2 was replaced by the discussion in the Bases of ITS Section 3.6, 11 Containment. 11
- 2. The discussion of the Primary Coolant System Design Pressure and Temperature in CTS Section 5.3.1 was replaced by the discussion in the Bases of ITS Section 3. 4, 11 Primary Cool ant System."
- 3. *The discussion of the Emergency Core Cooling System in CTS Section 5.3.3 was replaced by the discussion.in the Bases of ITS Section 3.5, 11 Emergency Core Cooling System. 11
- 4. The CTS Design Features description of spent fuel storag~ facilities
*contained both operatfng limitations and surveillance requirements.
Therefore, the requirements for f~el storage in the Region II fuel racks in CTS Section 5.4.2, was replaced by ITS 3.7.16, "Spent Fuel Assembly Storage." CTS Table 5.4-1 was also moved to that Specification and Figure 5.4-1 was moved to its Bases. The CTS requirements for Spent Fuel Pool boron concentration were moved to ITS 3.7.15, "Spent Fuel Pool Bdron
- Concentration."
5
INTRODUCTION: CHAPTER 4.0. DESIGN FEATURES G. THE MAJOR DIFFERENCES BETWEEN ITS AND ISTS
- 1. The ISTS 4.3.1.1 requirements for the Spent Fuel Storage Racks are identified.
separately for Region I (ITS 4.3.1.1) and Region II (ITS 4.3.1.2) of the storage racks .
- 6
ATTACHMENT 1 PALISADES NUCLEAR PLANT
- CHAPTER 4.0 - DESIGN FEATURES PROPOSED TECHNICAL SPECIFICATIONS
Design Features 4~0
- 4.0 DESIGN FEATURES 4.*1 Site Location The Palisades Nuclear Plant is located bn property owned by Consumers Energy on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city limits of South Haven, Michigan.
The minimum distance to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 677 meters. 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor core shall contain 204 fuel assemblies. Each assembly shall consist of a matrix of zircaloy-4 clad fuel rods with an initial composition of depleted, natural, or slightly enriched uranium dioxide (U0 2 ) as fuel materi~l. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution. Poison may be placed in the fuel bundles for*long-term reactivity control. 4.2.2 Control Rod Assemblies The r~actor ~cire shall ~ont~i~ 45 control rods. Four of these control rods may consist of part-length absorbers. The control material shall be silver-indium-cadmium, as approved by the NRC .
- Palisades Nuclear Plant 4.0-1 Amendment No. 01/20/98
Design Features 4.0
- 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The Region I fuel storage r~cks (See Figure B 3.7.16-1) are designed and shall be maintained with:
- a. Fuel assemblies having a maximum enrichment of 4.40 weight percent; br K~f ~ 0.95 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section.9.11 of the FSAR.
- c. A nominal 10.25 inch center to center distance between fuel assemblies with the exception of the single Type E rack which has a nominal 11.25 inch center to center distance between fuel assemblies; arid
- d. New 6r partially spent fuel assemblies *. Assemblies with enri c.hments . _above. 3. 27 *weight percent U235 must contain 216 rods _which are either U02 , Gd 203 -U0 2 . or solid metal.
4.3.1.2 The Region II fuel storage racks (See Figure B 3.7.16-1) are designed and shall be maintained with;
- a. Fuel assemblies having maximum enrichment of 3.27
----- 'wefght -percertf; * * "!.
b~ Keff. ~ 0.95 if 'fully flooded with unborated water,. which includes allowance for uncertainties as described in Section 9.11 of the FSAR.
- c. A nominal 9.17 inch center to center distance between fuel assemblies; and
- d. Partially spent fuel assembli~s which meet the discharge burnup requirements of Table 3.7.16-1 .
- Palisades Nuclear Plant 4.0-2 Amendment No. 01/20/98
Design Features 4.0
- 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.l Criticality (continued) 4.3.1.3 The new fuel storage racks are designed and shall be maintained with:
- a. Fuel assemblies having a maximum average planar U235 enrichment of 4.20 weight percent;
- b. Keff ~ 0. 95 when flooded with either full density or
.low density (optimum moderation) water including allowances fdr uncertainties as described in Section 9.11 of the FSAR. *
- c. Assemblies must contain 216 rods which are either U02 , Gd 2 03 -U02 or solid metal.
- d. The pitch of the new fuel storage rack lattice being ~ 9.375 inches and every other position in
- the lattice being permanently occupied by an 8 x 811 structural steel or core plugs, resulting 11 in a nominal 13.26 inch center to center distance between fuel assemblies placed in alternating storage locations.
4.3.2 Drainage The spent fuel storage pool cooling system suction and discharge piping is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 644 ft 5 inches. 4.3.3 Capacity The spent fuel storage pool and north tilt pit are designed and shall be maintained with a storage capacity limited to no more than 892 fuel assemblies .
- Palisades Nuclear Plant 4.0-3 Amendment No. 01/20/98
ATTACHMENT 2 PALISADES NUCLEAR PLANT
- CHAPTER 4.0 - DESIGN FEATURES PROPOSED BASES (N/Afor Chapter 4)
ATTACHMENT3 PALISADES NUCLEAR PLANT CHAPTER 4.0 - *nESIGN FEATURES
- CTSMARKUP AND
-DISCUSSION OF CHANGES
- The Pa 1i sad e \!.r.:c.:.c:..;t~--=~!..c::>
Energy Company on the and one-half miles south Michigan. to the boundary be 677 meters. by Consumers 5.2 5.2.1 The co ainment structure completely ncloses the primary *coolant sys to minimize release of radi ctive material to the e ironment should a failure of e primary coolant system occur he prestressed, post-tensio concrete structure provides adequate biological shield g for both normal operation an accident situations and *s designed for low leakage at a esign pressure of 55 psig a 283'F. The principal de gn basis for the structure is at it be capable of withstandi the internal pressure resulti from a design basis loss-of-coo nt accident. In this event, t total energy containe n the water of the primary co ant system is assumed to be rel sed into the containment thro a double-ended break of the argest primary coolant pipe co cident with a loss of normal standby electrical power. S sequent pressure behavior is determined by the engineered s ety features and the combined influence of energy sources nd heat sinks *
- The external design pr sure of the containment shell i This value is appro ately 0.5 psig greater than the aximum external pressur psig.
at could be developed if the c ainment were sealed during period of low barometric pressur and high temperature d, sub~~~ti~ritly, th~~containmen atmosphere were cooled w a concurrent major rise in bar etric pressure. Vacuum breake are therefore not provided.
- c. I structure.
~\
Amendment No. Si,176
- 5.2 5.2.2
- a. All pe trations through the steel- ned concrete structure for elec ical conductors, pipe, ducts air locks and doors are of he dou e-barr1er design.
- b. e automatically actuated co ainment isolation valves ar designed to close upon high ad1ation or high pressure i containment structure. No ingle component failure in actuation syste* will pr tnt the isolation valves fr as designed.
cooling syst111 includes fo separate self-contained units ich cool the contain111nt ai during normal operation and *it the pressure rise.int~ event of a design accident. Thee units, with a total coolt water flow of 5580 gpm with an inl temperature of as*F, will r ve 230 x io* Btu/hr of hut.
- b. The con ain111ent spray syst111 1s capa e of re110ving 233 x 108 Btu/h (two pumps) frOll the contai nt atmosphere at 283°F by spr ing the wat~ frOll the 270,0 -gallon SIRW tank.
Re rculation of spray water fr the conta1n.. nt sump through h t e changers into the contain..n ataosphere 1s also provided. U ij1r his mode of operation, the at r1110val capability is 167 x
- Btu/hr based upon 4000 gpa component cooling water flow 'th 114°F inlet temperature th ough tht heat exchanger and 142 gpm of spray water flow at 283°E i~let temp1ratur1.
5.3.l
- a. In a cordance with the Cod requirements specifie in Se ion 4.2 of the FSAR w h allowance for no~
d radation pursuant to e surYtillance requir b. For 1 tll!C)trature o 6SO*F, except the prt be 1oo*F, and With a volu.e o approxiaately 10,900
,I 5-2 Amendment No. ~' ~' 166 May 22, 1995
- (Con)'1nued) te a r gnt circu ar. cy inder with 136 inches and an active height of c.
L;.-:+-r.J s~Ll.J..*.J.,.J,..,_r ot -e;,.c..o"':.__ 4..llD'."\ olf" r.f..,:~l-eH s+J
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£.o Amendment Ho. ~. 49, ~. ~. 171 Apr i 1 5 , 19 96
- FUEL STORAGE New Fyel Storage
- a. The pitch of the new fuel storage rack lattice is ~ 9.375 inches, and every other position in the lattice shall be permanently occupied by an s* x a* structural steel box beam or core plugs such that the minimum center-to-center spacing of new fuel assemblies 1n the alternating storage array is 13.26". This distance in the alternating storage lattice is sufficient so that K," will not exceed 0.95 where fuel assemblies with 216 U0 2 or Gd 203 *U0 2 fuel rods or metal rods and a maximum average planar enrichment in the U0 2 or Gd 203 *U0 2 fuel rods of 4.20 w/o U231 are in place and optimum moderation is assumed. The calculated K includes a ro riate conser~atisms as described in iemens Nuc ear Power Corporation b.
- c. storage racks ar
- S*4 Amendment No. H, 49, ~; 140 January 23, 1992
- . C\\\
Lt)~
~ fyel Storage
- a. Irradi ted fuel bundles ~* J be stored, prior o off-site shipment
- tainless steel-1 'ned sent fuel lb. -/{Deleted}/
I I @
@@ The spent fuel storage pool and spare (north} tilt pit are divided into two regions identifie as Region I and Region II as illustrated in Figure - . Region I racks are designed and shall 4.3.llc,... be maintained with a nominal l0.2s* center-to-center distance between fuel assemblies with the exception of the single Type E rack which has a nominal 11.25* center-to-center distance between fuel assemblies. The Region I spent fuel storage racks are designed such that fuel having a maximum assembly planar average
'-\ ..3.1. It\. U231 enrichment of 4.40 w/o placed in the racks would result in a
--- Kl'ff equivalent to s 0.95 when flooded with unborattd water. The
'\.3.!.lh K... of s 0.95 includes a _conservative allowanc*e for uncertainties.
- 4. 3.1, f! For enrichments above 3.27 w/o U231 , the fuel assemblies must conhin 216 rods which are either U0 2 , Gd2') 3 *U0 2 or. sol id metal.
- je. 7 (De1etj() I @
~.Q s.1.1~-
lg. The sjent fuel rackf are designed u/4 Class I str}?cture.I § I h. / (Oelet!d>I @ Storage in Region II of the spent fuel pool and spare (north) tilt* ct.~- t 1J. . pit stiall be restricted bf burnup and enrichment 1imits specified ( See.Also 3, 7 7 in ttabli 5.Mu~-tr;.Y~ 3.1 . 11.,.. 1 1 ~ nti needed or f el storage, on*e Re ion II rack in th *B northeast corner f the spent fuel ol may be removed and L i!I. t 1 . re laced with t cask anti-ti in device. .
~r.:.r"tJ ':I:\S) @
- < A7JD 5-41 Amendment No. ~. i-t-t, 140 January 23, 1992
N ----J ...... R99ion II 11 x 11 11 x 11x 176.00" W-1 *-2 Ref. R99lon 11 1 x " R99ion 11 1 x 11 ..,.
.D an
'9 I
W-l W-4 r-= MAIN PDCn.. ("""' r-
&o.oo" Region 11 r IGURE S.4-1 6 )( 1 Ref. W-S LL-~.....L!::~~~~~~
SPEMT FUEL ARRANGEMENT I North Tl It Pit
~ 2s2.oo* Ref. -------1>1
TABLE 5.4-1 Fuel Burnup Requirements Storage in Region II of the Spent Fuel Pit Initial Discharge Burnup w/o GWO/MJ 1.5 0 1.9 5.2 8.5 11.5 14.1
,(see J.7>-
16.6 2.8 18.8 3.0 3.2
. 3.27 Linear interpolation between two consecutive point yield conservative results.
Amendment No. 105 July 24, 1987
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- ADMINISTRATIVE CHANGES (A)
ATTEiCHMENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN-FEATURES A.1 All reformatting and renumbering are in accordance with NUREG-1432. As a result, . the Technical Specifications (TS) should be more readily readable, and therefore understandable by plant operators as well as other users. The reformatting, renumbering, and rewording process involve no technical changes to existing Technica{ Specifications. Editorial rewording (either adding or deleting) is made consistent with NUREG-1432. During Improved Technical Specification (ITS) development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or implied) to the TS: Additional information has also been added to more fully describe each subsection. This wording is consistent with NUREG-1432. Since the design is already approved by the NRC, adding more details does not result in a technical change. A.2 A sentence is added to the CTS 5.3.2d discussion regarding the material contained in the control rods. The proposed sentence states "The control material shall be silver-indium-cadmium as approved by the NRC." This is the current material contained in the Palisades control rods and is discussed in the FSAR. Therefore, it imposes no new requirements and is therefore considered to be an administrative change .. This change
*is consistent with NUREG-1432.
A.3 Wording is added to CTS 5.3.2b to discuss substituting zirconium alloy or stainless steel filler rods for fuel rods in accordance with NRC staff approved codes and
*methods. The proposed wording states "Limited substitutions of zirconium alloy or stainless steel filler rods, in accordance with approved applications of fuel rod configurations may be used~ Fuel assemblies shall be limited to those fuel designs that .
have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed* representative testing may be placed in nonlimiting core regions. " As stated, any changes must be done with NRC staff approved codes and methods. Therefore, adding this statement only clarifies that these changes must be done with an NRC approved methodology. Guidance on this subject was provided jn NRC Generic Letter 90-02 and a subsequent Supplement 1 which provided staff clarification on this issue. This is considered to be an administrative change since the wording is clarifying the NRC expectations and requirements. This wording is consistent with NUREG-1432 .
- Palisades Nuclear Plant Page 1of6 01/20/98
- A.4 ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN"FEATURES CTS 5 .1 states "The Palisades reactor shalf be located on .... " The proposed ITS states "The Palisades Nuclear Plant is located in .... " This change more appropriately reflects the plant name. In addition "shall be" is changed to "is" to improve the sentence wording-and reflect the first that the Palisades Nuclear Plant has already been built.
These changes are considered to be administrative changes since no requirements have changed. A.5 The Palisades CTS does not address inadvertent draining of the storage pool. The - suction and discharge piping of the cooling system for the storage pool was designed to prevent inadvertent draining. The discharge piping is at 647' and contain8 a siphon breaker. The bottom of the suction piping is at elevation 644' 5." Since these piping arrangements are permanent plant features, and no additional operational requirements have been imposed the inclusion of this information fato the proposed ITS is considered to be an administrative change. This change is consistent with the intent of NUREG-1432. A.6 CTS 5 .4. la contains certain design aspects of the new fuel storage racks and includes a reference to Siemens Nuclear Power Corporation Report EMF-91-1421 (NP) for the appropriate conservatism used in the calculation of Keff* In proposed ITS 4. 3 .1. 3, reference to the Siemens Nuclear Power Corporation Report has been replaced by a reference to FSAR Section *9 .11. _Section 9 .11 of the FSAR documents the design and analysis for the Fuel Handling and Storage Systems. This change is considered administrative in nature since it does not alter the design or analysis assumptions of the new fuel storage racks, but merely revises the reference of the document which contains the uncertainiie.s used in the determination of Keff*: This change -is consistent with NUREG-1432. TECHNICAL CHANGES - MORE RESTRICTIVE (M) There were no "More Restrictive" changes added to this chapter .
- Palisades Nuclear Plant Page 2of6 01/20/98
- ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN.FEATURES LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
LA.1 CTS Specification 5 .1 specifies that "The Palisades reactor shall be located on 437 acres owned by Consumers Power Company ... " This is changed in the proposed ITS to state "The Palisades plant is located on property owned by Consumers Energy ... " .,. The specific amount of land will be located in the FSAR. In addition, this section also
*states "Figure 5-1 shows the plan of the site." This figure will also be moved from the TS ~nd will reside in the FSAR. The minimum distance to the boundary of the exclusion area which is the primary factor in determining offsite doses is still retained in the proposed ITS. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removal of the specific acreage owned by Consumers Power and Figure 5-1 from the CTS and placing it in the FSAR will not adversely impact safety. Changes to the information in the FSAR will be made in accordance with the provisions of 10 CFR 50.59. The removal of this information maintains consistency with NUREG-1432.
LA2 Descriptive information q:mtained in CTS Specification 5 .2, Containment Design Features, is not included in the proposed ITS Chapter 4.0, Design Features, as this type of design information is contained in the FSAR. Therefore*CTS 5.2.1, Containment Structure, CTS 5.2.2, Penetrations, and CTS 5.2.3, Containment Structure Cooling Systems are removed from the TS and this type of information will be addressed in the FSAR. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removing this information from the CTS and controlling it under this FSAR does not adversely impact safety. Changes to the information in the FSAR will be made in accordance with the provisions of 10 CFR 50.59. The removal of this information maintains consistency with NUREG-1432. LA. 3 Descriptive information contained in CTS Specification 5. 3 .1, Primary Coolant System Design Pressure and Temperature, is not included in the proposed ITS Chapter 4.0, Design Features, as this type of information is currently contained in the FSAR. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Changes to the information in the FSAR will be made in accordance with the provisions of 10 CFR 50.59. Removing this information from the CTS and controlling it under the FSAR does not adversely impact safety. The removal of this information maintains consistency with NUREG-1432 .
- Palisades Nuclear Plant Page 3of6 01/20/98
- ATTACI'ENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN-'FEATURES LA.4 Descriptive information contained in CTS Specification 5.3.2, Reactor Core and
. Control, part "a," "c," and "d," is not included in the proposed ITS Chapter 4.0, Design Features, as this type of information is currently contained in the FSAR. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Changes to the information in the FSAR will be made in accordanc~ with the provisions of 10 CFR 50.59. Removing this information from the CTS and controlling it under the FSAR does not adversely impact safety. The removal of this information maintains consistency with NUREG-1432.
LA.5 . Descriptive information contained in CTS Specification 5.. 3.2b discussing the number of fuel rods, "approximately 43,000," and the "sintered" U0 2 "pellets," is not included in the proposed ITS Chapter 4.0, Design Features. This type of information is contained in or, in the case of the number of fuel rods can be inferred from the information presented in the FSAR on the number of assemblies in the reactor and number of rods in an assembly. This information is not required by 10 CFR 50.36(c)(4) to be in the TS.* Removing this informationfrom the CTS and controlling it under the FSAR does not adversely impact safety. Changes to the . information in the FSAR will be inade in accordance with the provisions of 10 CFR 50.59. The removal of this information maintains consistency with NUREG-1432. LA.6 . Descriptive information contained in CTS Specification 5.3.3, Emergency Core Cooling System, is not included in the propose4 ITS Chapter 4.0, Design Features, as this type of design information is contained in the FSAR. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removing this information from the TS and controlling it under the FSAR does not adversely impact safety. Changes to the information in the FSAR will be made in accordance with the provisions of 10 CFR 50.59. The removal of this information maintains consistency with. NUREG-1432. LA.7 CTS Spedfication 5.4.lb contains information which is not included in the proposed ITS Chapter 4.0, Design Features. CTS 504. lb states that "New fuel may also be stored in *shipping containers." This is a statement which is not relevant to the discussion of the new fuel storage racks and does not place any restrictions on fuel storage. The proposed ITS discusses the design of the fuel storage facilities rather than the shipping facilities. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removing this information from the CTS and placing it under licensee control does not adversely impact safety. The removal of this information maintains consistency with NUREG-1432 .
- Palisades Nuclear Plant Page 4of6 01/20/98
- _ATTACHMENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN*FEATURES LA.8 CTS Specification 5. l .4c contains information which is not included in the proposed ITS Chapter 4.0, Design Features. CTS 5.4.lc states "The new fuel storage racks are designed as a Class 1 structure." This information is currently contained in the FSAR.
This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removing this. information from the CTS and controlling it in the FSAR does not adversely impact safety. Changes to the information in the FSAR will be made in accordance with the provisions of 10 CFR 50.59. The removal of this information maintains consistency with NUREG-1432. . LA.9 CTS 5.4.2a contains information which is descriptive in nature and is not included in the proposed ITS Chapter 4.0, Design Features. CTS 5.4.2a states "Irradiated fuel
- bundles will be stored, prior to off-site shipment in the stainless steel-lined spent fuel pool." This type of information is contained in refueling and fuel handling procedures which are under licensee control. This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removing this information from the CTS and controlling it in plant procedures does not adversely impact safety. The removal of this information maintains consistency with NUREG-1432 .
- LA.10 CTS 5.4.2g contains design related information which is not included in the proposed ITS Chapter 4.0, Design Features. CTS 5.4.2g states "The spent fuel racks are designed as a Class 1 structure." This information is currently contained in the FSAR.
This information is not required by 10 CFR 50.36(c)(4) to be in the TS. Removing this information from the CTS and controlling it in the FSAR does not adversely impact safety. Changes to the information in the FSAR will be made in accordance with the provisions of 10 CFR 50.59. The removal of this information maintains* consistency with NUREG-1432. LA.11 CTS 5.4.2 contains a NOTE which states "Until needed for fuel storage, one Region II rack in the northeast comer of the spent fuel pool may be removed and replaced with the cask anti-tipping device." This information is currently contained in the FSAR and is pot included in the proposed ITS Chapter 4.0, Design Features. Removing this information from the CTS and controlling it in the FSAR does not adversely impact safety. This inforination is not required by 10 CFR 50.36(c)(4) to be in the TS. Changes to the information in the FSAR will be made in accordance with the
. provisions of 10 CFR 50.59. The removal of this information maintains consistency with NUREG-1432 .
- Palisades Nuclear Plant Page 5of6 01/20/98
- ATTliC1'ENT 3 DISCUSSION OF CHANGES CHAPTER 4.0, DESIGN-FEATURES LA.12 CTS Figure 5 .4 provides a graphical representation of the spent fuel storage facilities including the Main Pool and the North Tilt Pool. Region I and II are shown which represent acceptable areas for fuel assembly storage based on discharge burnup. This figure will be moved to the TS Bases for 3. 7 .16, Spent Fuel Storage. This information is not required by 10 CPR 50.36(c)(4) to be in the TS. Removing this information from the CTS and controlling it in the Bases will not adversely impact safety. Changes,.
to the information in the Bases will be made in accordance with the TS Bases Change Control Program which is discussed in proposed Specification 5.5.12. The removal of this information maintains consistency with NUREG-1432. LESS RESTRICTIVE CHANGES (L) There were no "Less Restrictive" changes made to this chapter. RELOCATED (R)
- There were no "Relocated" changes made to this chapter .
- Palisades Nuclear Plant Page 6of6 01/20/98
. ATTACHMENT 4 PALISADES NUCLEAR PLANT
- CHAPTER 4.0 - DESIGN FEATURES NO SIGNIFICANT HAZARDS CONSIDERATION
- ADMINISTRATIVE CHANGES (A)
NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 CHAPTER 4.0, DESIGN. FEATURES The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering* Plants." Some of the proposed changes involve reformatting, renumbering, and rewording of Technical Specifications. These changes, since they do not involve technical changes to the Technical Specifications, are administrative.
- This type of change is connected with the movement of requirements within the current requirements, or with the modification of wording which does not affect the technical content of the current Technical Specifications. These changes will also include nontechnical modifications of requirements to conform to the Writer's Guide or provide consistency with the Improved Standard Technical Specifications in NUREG-1432. Administrative changes are not intended to add, delete, or relocate any technical requirements of the current Technical Specifications.
In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not
- represent a significant hazards consideration. The following is provided in support of this conclusion.
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specification. These modifications involve no technical changes to the existing Technical Specifications. The majority of changes were done in order to be consistent with NUREG-1432. During the development of NUREG-1432, certain wording preferences or English language conventions were adopted. The changes are administrative in nature and do.not impact initiators of analyzed events. They also do not impact the assumed mitigation of accidents or transient events. Therefore, the changes do not involve a significant increase in the probability .or consequences of an accident previously evaluated .
- Palisades Nuclear Plant Page 1of6 01/20/98
- 2.
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 4.0, DESIGN FEATURES Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The proposed changes involve reformatting, renumbering, and rewording of the existing Technical Specifications. The changes do not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The changes will not impose any new or different requirements or eliminate any existing requirements. Therefore, the changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in margin of safety?
The proposed changes involve reformatting, renumbering, and rewording of the existing Technical SpeCifications. The changes are administrative in nature and will not involve any technical changes. The changes will not reduce a margin of safety because it has no impact on any safety analysis assumptions. Also, since these changes are administrative in nature, no question of safety is involved. Therefore,
- the changes do not involve a significant reduction in a margin of safety .
- Palisades Nuclear Plant Page 2 of 6 01/20/98
- ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 4.0, DESIGN"FEATURES MORE RESTRICTIVE CHANGES (M)
There were no-More Restrictive" changes made in Chapter 4.0 .
- Palisades Nuclear Plant Page 3 of 6 01/20/98
- NO SIGNIFICANT HAZARDS CONSIDERATION ATTACHMENT 4 CHAPTER 4.0, DESIGN FEATURES LESS RESTRICTIVE CHANGES - REMOVAL OF DETAILS TO LICENSEE CONTROLLED DOCUMENTS (LA)
The Palisades Nuclear Plant is converting to the Improved Technical Specifications (ITS) as outlined in NUREG-1432, "Standard Technical Specifications, Combustion Engineering Plants." Some of the proposed changes involve moving details (engineering, procedural, etc.) out of the Technical Specifications and into a licensee controlled document. This information may be moved to the ITS Bases, FSAR, plant procedures or other programs controlled by the licensee. The removal of this information is considered to be less restrictive because it is no longer controlled by the Technical Specification change process. Typically, the information moved is descriptive in nature and its removal conforms with NUREG-1432 for format and content. In accordance with the criteria set forth in 10 CFR 50.92, Palisades Nuclear Plant staff has evaluated these proposed Technical Specification changes and determined they do not represent a significant hazards consideration. The following is provided in support of this conclusion .
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
Analyzed events are assumed to be initiated by the failure of plant structures, systems or components. Consequences of a previously analyzed event are dependent on the initial conditions assumed for the analysis, and the availability and successful functioning of the equipment assumed to operate in response to the analyzed event. The proposed changes move details from the Technical Specifications to a licensee controlled document. The removal of details from the Technical Specifications is not assumed to be an initiator of any analyzed event. The proposed changes do not reduce the functional requirement or alter the intent of any specification. As such, the consequences of an accident remain unchanged. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated .
- Palisades Nuclear Plant Page 4 of 6 01/20/98
- 2.
ATTACHMENT 4
.NO SIGNIFICANT HAZARDS CONSIDERATION CHAPTER 4.0, DESIGN FEATURES Does the change create the possibility. of a new or different kind of accident from any accident previously evaluated?
The proposed changes move detail from the Technical Specifications to a licensee controlled document. The changes will not alter the plant configuration (no new or different type of equipment will. be installed) or make changes in methods governing ,. normal plant operation. The changes will not impose different requirements, and adequate control* of information will be maintained. The changes will not alter assumptions made in the safety analysis and licensing basis. Therefore, the changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does this change involve a significant reduction in a margin of safety?
Margin of safety is determined by the design and qualification of the plant equipment, the operation of the plant within analyzed limits, and the point at which protective or mitigative actions are initiated. There are no design changes or equipment performance parameter changes associated with this change. No setpoints are
- affected, and no change is being proposed in the plant operational limits as a result of this change. The proposed changes remove details from the Technical Specifications and place them under licensee control. Removal of these details is acceptable since this information is not directly pertinent to the actual requirement and does not alter the intent of the requirement. Since these details are not necessary to adequately describe the actual regulatory requirement, they can be moved to licensee controlled document without a significant impact on safety. Therefore, the proposed changes do not involve a significant reduction in a margin of safety .
- Palisades Nuclear Plant Page 5 of 6 01/20/98
- LESS RESTRICTIVE CHANGES (L)
ATTACHMENT 4 NO SIGNIFICANT HAZARDS CONSIDERATION
. CHAPTER 4.0, DESIGN FEATURES There were no--~'Less Restrictive Changes" made in Chapter 4.
i I I
- Palisades Nuclear Plant Page 6 of 6 01120/98
ATIACHMENT 5 PALISADES NUCLEAR PLANT
- CHAPTER 4.0 - DESIGN- FEATURES MARKUP OF NUREG-1432 TECHNICAL SPECIFICATIONS AND BASES
, Design Features 4.0 .
- 4.0 DESIGN FEATURES 4.1 Slte Locition 1t'frext #scription/6f the site }§cation!.¥
\....<;r:rJS£KT) j (!)
4.2 Reactor Core 4.2.1 Eyel Assemblies The ructor shall contain assembly fl\ shall consist of a matrix o fuel rods with \:._} in initiil composition ott'natur1ld)Or slightly enriched urinium *I~ dioxide (U0 2 ) iS fuel material. [i*ited substitutions of zirconium illoy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test
- assemblies that have not completed representative testing may be placed in non11miting core regions *
- ea ...-t,.. .. \ ,..,.,ls Ma. 1 Fuel Storage C.o.-s;st ~+- F,..+- l~.a.. 4.3.l Criticality See F; '"'"...'- ~ 3.1. "-()
a..bs""'b~r!: are designed and shall 4.3.1.l
- 1. Fuel assemblies having a maximum U-235 enrichment of S'we1ght percent; (!) I
- b. . kett ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in f5ect1on.9~of thi.FSARf;" J© (continued)
CEOG STS 4.0-1 Rev l, 04/07/95
- CHAPTER4.0 INSERT The Palisades Nuclear Plant is located on property owned by Consumers Energy on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city limits of South Haven, Michigan. The minimum distance to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 677 meters .
- 4.0-1
Design Features 4.0
- 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)
(c. See ~3.1.l. uJ :~ ~ e.~c.qll~ I)~ ~e. '5.:":J'~ ~l'e. £
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c.e~-h?,. i:b*+o.."'CP- be..+.,.l~e."'
.f..-..l. .::;..stie.-b\.:-c.t j J @)
Ass~""' \.\:<-s w:~ ..1\,,.:J.- -~...ts [f. New or partially spent fuel assemblies with a
~\.o.,.t.... 3.'2..1 .,..i/o UzJ~ ,.,..wST discharge burnup in the "unacceptable range" of s-. l\.l.L?...
Figure (3.7.17-1] will be stored in compliance with C...1>>"'+.:..~.... 2-\IA '('oJ..5 "'1\....:-~ o..~ the NRC approved [specified document containing the analytical methods, title, date, or specific
.~i ~ \..\Oz.. 1 C, J.z. 01 - U ()L o-r configuration or figure].]
s.~t*~ ..... e.-t.JL. ,.3.1.~ The new fuel storage racks are designed and shall be
-@ miflntained *with:'
<J:tJSE:KT '13.t.'2..
a. uh~ ~lo..J~ ...,;.\(....:- b.
~;--<<-.... .{:._\\ <l~"'s:+-\ """
IO.,..> ot~.,_;}l L~---
- c. k.,. ~ . 98 _if moder ed by aqueo_u foam_, w c 1"\QJ~-°""'J IJo...~r inclu es an allow ce for uncer inties as des tbed in [Se ion _9.1 oft FSAR]; and
@@. A nominal (10] inch center to center distance between fuel assembl ie~ pl aced in the storage...--
racks.
@ Ass~ U;es t"\...s+ c.o"'-l-o.:~ UI,,, .,,. .. Js i,;ih:""' ~ ~i.J.l-... tE'\
4.3.2 Drainage lAOi_, 6cl?...~-U.oL ov- ~.hl .... e+~. 0 The spent fuel storage pool 1s designed and shall be main~~ to prevent inadvertent draining of the pool below elev~t\~n ~*1 [j) coo; ... ~ SIM:.' '~"' ~"' ; "'~' il;t;"'j 1-j'i ~
. \!_) (continued)
CEOG STS 4.0-2 Rev 1, 04/07 /95 Tk- f~~ Q,f" ~ --~~ .{:~ <i-h>r'-"\ ~ .,. c..c..~ \a..-"-.""-:- be.:.--) ~ q, ~1 S l~c...""-~5 o...J. ~.,..___'\ o~ pas:ko...._ l"- ~ la..~.'~ be..' . . . j f-'4-~"'<!"'-1-1\ 0 C.C.."'-f \el, b'\ a...... g** 8"
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Design Features 4.0
- 4.0 DESIGN FEATURES 4.1 Sitt Location [Text description of the site location.].
4.2 Reactor Core 4.2.1 fyel Assemblies The reactor shall contain [217] fuel assemblies. Each assembly shall consist of a matrix of [Zircalloy or ZIRLO] fuel rods with an initial composition of natural or slightly enriched uranium ,_ .Se.e dioxide (U0 2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. 4.2.2 [Control Bodl Assemblies The reactor core shall contain [91] control element assemblies {CEAs). The control material shall be [silver indium cadmium, boron carbide, or hafnium metal] as approved by .the NRC.
- (riff£'(.\.,~:~~ lte~'*-"lt:'
preJ.'o ... s / 014.3.1.~ Tht s (See- k* _.,_ g ?.<.lb-I fuel storage rack art designed and shall be 6:J
,,,..1 e
- V (i) Hin a neCi with:
- a. Fuel assemblies having a maximwa U-235 enrichment 1.:
weight percent; . IQJ
- b. eH s 5 if fully flooded with unborated water, which includes an allowance for uncertainties as described in .{~~t1.~rn 9 .~ of the .FSA~ / If'
*@ tJJ (continued)
CEOG STS 4.0-1 Rev 1, 04/07/95
Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) (c'. ensity [d. A nominal [10.4] inch center to center distance between fuel assemblies placed in [the low density .. fuel storage racks];] il'seq_ ~J.L\) [e. New or partially spent fuel assemblies with a (9 discharge burnup in the *acceptable range* of Figure [3.7.17*1] may be allowed unrestricted storage in [either] fuel storage rack(s}; and] c.:rs S"",4.1.l, s.~.2tA (.,.J h: c.k M~C..;- .fl-..L a~s°'c.. ~c.. bv-J'P\V-! i.-------"1 t'-1!.6~"<"""1!!_.l-.s '°
~1.\~ ?..1. \b-1 .
4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
- a.
b. Fuel assemblies having a maximUll U-235 enrichment of [4.5] weight percent; k.tt ~ 0.95 if fully flooded with unborated water, which includes an allowance for uncerta;nties as described in [Section 9.1 of the FSAR];
- c. k.tt s 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR]; and
- d. A nominal [10] inch center to center distance between fuel assemblies placed in the storage racks.
4.3.2 Drainage The spent fuel storage pool ts designed and shall be maintained to prevent inadvertent draining of the pool below elevation [23 ft]. (continued} CEOG STS 4.0-2 Rev 1, 04/07/95
Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3.3 Capacity
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CEOG STS 4.0-3 Rev 1, 04/07 /95
ATTACHMENT 6 PALISADES NUCLEAR PLANT
- CHAPTER 4.0 - DESIGN FEATURES JUSTIF'ICATION FOR DEVIATIONS FROM NUREG-1432
- TECHNICAL SPECIFICATIONS ATTACHMENT 6 JUSTIFICATIONS FOR DEVIATIONS CHAPTER 4.0, DESIGN FEATURES NOTE: The first five justifications for these changes from NUREG-1432 were generically used throughout the individual LCO section markups. Not all generic justifications_ are used in each section.
- 1. The brackets have been removed and the proper plant specific information or value has been provided.
- 2. Editorial change for clarity or for consistency with the Improved Technical Specifications (ITS) Writer's Guide.
- 3. The requirement/statement has been deleted since it is not applicable to this facility.
The following requirements have been renumbered, where applicable, to reflect this deletion.
- 4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the facility specific nomenclature, number, reference, system description, or analysis description.
- 5. This change reflects the current licensing basis/technical specifications.
- 6. The proposed Palisades ITS is structured slightly different from NUREG-1432 to better present the Palisades spent fuel pool design and storage_ restrictions. Section 4.3.1.1 in NUREG-1432 has been divided into two parts with 4.3.1.1 addressed the Region I storage restriction8 and 4.3.1.2 addressing the Region II storage restrictions.
This results in a better understanding of the requirements associated with each Region.
- 7. In proposed ITS 4.3.2, Drainage, the phrase "The spent fuel storage pool" is revised to state "The spent fuel storage pool cooling suction and discharge piping ... " This deviation reflects the design features of the Palisades storage pool which prevent inadvertent draining .
- Palisades Nuclear Plant Page 1of1 01/20/98}}