ML18093B378

Proposed Tech Specs,Redefining Fully Withdrawn to Allow Positioning of Rod Control Cluster Assemblies in Band Between 222 & 228 Steps When Fully Withdrawn from Reactor

ML18093B378
Person / Time
Site:	Salem
Issue date:	01/03/1989
From:	Public Service Enterprise Group
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ML18093B377	List: ML18093B376 ML18093B378
References
NUDOCS 8901100247
Download: ML18093B378 (55)

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ATTACHMENT 1 LCR 88.-15 Description of Change Revise the Salem Unit 1 and Unit 2 Technical Specifications as follows:

1. Definitions - add a definition for the fully withdrawn position of the Rod Cluster Control Assemblies (RCCAs).

2. Modify definition 1.28, SHUTDOWN MARGIN, Specifications 3.1.3.4 and 3.10.1, and Bases 2.1.1 and 3/4.1.3 to incorporate the new definition of "FULLY-WITHDRAWN".

3. Replace Figure 3.1-1 to incorporate the new definition of FULLY WITHDRAWN.

4. Delete Figure 3.1.2 from Unit 1.

5. Modify Specification 3.1.3.3 to clarify rod drop test requirements.

6. Modify Unit 2 Specification 3.1.3.2.2 to incorporate the rod drop testing requirements previously in Specification 3.10.5.

7. Add to Unit 1 Specification 3.1.3.2.2 rod drop test requirements as included in Unit 2 to achieve consistency between units .

8. Delete Unit 2 Specification 3.10.5.

Reason for Change Proposed revision items 1 through 3 are being requested to address potential rod.wear concerns as seen previously at other Westinghouse plants. These items redefine FULLY WITHDRAWN to be between 222 and 228 steps withdrawn. This request is similar to requests made by and approved for the Callaway, Trojan arid Point Beach plants.

Proposed revision Item 4 is being requested to delete the curve implementing three loop operations which is not currently allowed but is still affected by redefining FULLY WITHDRAWN. Rather than modifying this specification, it is being deleted. This is consistent with the Unit 2 specifications.

Item 5 is being requested to clarify that rod drop test times are to be performed from 228 steps withdrawn. With the proposed redefinition of FULLY WITHDRAWN, test times could be performed from 222 steps withdrawn if this clarification was not made.

Proposed revisions 6 through 8 are being requested to correct an inconsistency present in the current Unit 2 Technical Specifications. Previously, a change was approved that no longer required that the Analog Rod Position Indication (ARP!) be

LCR 88-15 operable in Modes 3, 4 and 5. This eliminates the need for

• Specification 3.10.5 since the other requirements are being addressed in specification 3.1.3.2.2. The rod drop test requirements are being added to Unit 1 for consistency between units.

Justification for Change

1. Items 1 through 5 NRC Information Notice No. 87-19 warns of potentially significant safety problems that can result from the perforation and cracking of RCCAs in Westinghouse PWRs. NRC AEOD/E613 documents actual cases of RCCA degradation at Point Beach Unit 2, Kewaunee, and Haddam Neck due to control rod wear. The purpose of the proposed Technical Specification changes is to allow repositioning of the "parked" position of the RCCAs. This repositioning of surfaces can effectively and safely increase the current RCCA lifetime in the Salem plants.

This LCR defines FULLY WITHDRAWN to mean that the RCCAs can be positioned within the range of 222 to 228 steps withdrawn. The specific definition of the fully withdrawn position for each cycle will be defined in the Reload Safety Analysis for that cycle. The definition will be implemented by an administrative procedure which will be controlled by Section 6. 5 .1 of the Technical Specifications. *The procedure will assure that the specific fully withdrawn position defined for the cycle has been considered in the cycle specific safety evaluation.

Item 4 deletes Unit 1 Figure 3.1-2, three loop operation, which is not an approved mode of operation at this time.

This also achieves consistency with the Unit 2 specifications.

Item 5 - With the redefinition of FULLY WITHDRAWN, it is necessary to clarify the point from which rod drop time tests are initiated. Analyses assumed that rod drop times are performed from 228 steps withdrawn. With the proposed redefinition of FULLY WITHDRAWN, rod drops could be performed from 222 to 228 steps withdrawn. This change will preclude potential violations of the safety analysis by ensuring testing is performed from a conservative starting point.

2. Items 6 through 8 The bases of special test exception 3.10.5 in the Unit 2 specifications states that the "exception is required since

• the data necessary to determine rod drop time is derived from the induced voltage in position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the position indication systems remain OPERABLE." LCR 85-12

,J, LCR 88-15 previously eliminated the requirement for having the ARPI system operable during Modes 3, 4 and 5 and allows use of the group demand position indicators for rod position indication for these modes. Section 3.10.5 did not reflect the previous change from the ARPis to the group demand counters. As currently written, 3.10.5 refers to equipment that is not required to be applicable. For this reason, the special test exception 3.10.5 is no longer necessary. The requirements of the special test exception which allow only one rod bank to be withdrawn at a time and restricts the core Keff to be less than or equal to 0.95 were originally put into 3.10.5 to preclude an inadvertent criticality.

These requirements continue to be appropriate during the performance of the rod testing during rod position indication calibration. For this reason the above requirements are being moved into Specification 3.1.3.2.2 for Unit 2 and added to Unit 1.

Significant Hazards Evaluation PSE&G has reviewed the proposed changes in this LCR and has determined that they do not constitute a significant hazards consideration as discussed below for each item.

1. Rods Fully Withdrawn Definition (Items 1-5)

A safety evaluation has been performed to address repositioning the fully withdrawn position of the RCCAs (Attachment 4). The evaluation considered the effects of the proposed technical specification changes on the following areas:

a. Small Break LOCA

b. Large Break LOCA

c. Short and Long Term LOCA

d. Steam Generator Tube Rupture

e. Post-LOCA Long Term Cooling

f. Hot Leg Switchover to Prevent Potential Boron Precipitation

g. Blowdown Reactor Vessel and Loop Forces

h. Non-LOCA Transients The conclusions of the evaluation are as follows:

a. The changes in the definition of the fully withdrawn RCCA position proposed create no significant changes in the affected safety parameters involved in verification of current technical specification limits. The involved safety parameters include those parameters normally addressed by the cycle specific Reload Safety Evaluation Checklist. The change of the fully withdrawn position from 228 steps to 222 steps or higher involves only a small amount of absorber

• being inserted into the active region of core and does not result in any design or regulatory limit being exceeded.

,J LCR 88-15

b. No FSAR safety limits are exceeded based on the proposed technical specification change. The position of the control and shutdown banks, relative to each other in the core will not change; therefore the limiting axial power distribution assumed for the DNB analyses remain applicable. The FSAR conclusion that the DNBR design basis acceptance criteria is met for the Condition II events remains valid.

Additionaliy, there is no significant impact on any core physics assumptions and design peaking factors important to the non-LOCA safety analyses and the reload verification.

c. The proposed change does not invalidate current control rod drop times or other tripped rod characteristics assumed in the LOCA licensing basis analysis.

Operation of the Salem Units in accordance with this proposed technical specification change:

a. Would not involve a significant increase in the probability or consequences of an accident previously evaluated for the Salem units, since the changes caused by repositioning the fully withdrawn position of the control rods are bounded by those assumed in the accident analyses.

b. Would not create the possibility of a new or different kind of accident from any accident previously evaluated for the Salem Units, since no plant hardware changes are required by this change.

c. Would not involve a significant reduction in a margin of safety, since the margin which was assumed in the accident analyses bounds the change proposed.

Conclusion Based on the above, we have concluded that the proposed technical specification changes correspond to example II.6 of guidance provided by the Commission in Federal Register FR14870 for Amendments Considered Not Likely to Involve Significant Hazards Consideration.

2. Elimination of Special Test Exemption 3.10.5 (Items 6-8)

Operation of the Salem Units in accordance with this proposed Technical Specification change:

a. Would not create a significant increase in the probability or consequences of an accident previously evaluated for the Salem Units since the change is administrative in that it eliminates an unnecessary specification and incorporates the requirements into an existing specification. Additionally, it imposes a

• like requirement into the Unit 1 Technical Specification;

i LCR 88-15

b. Would not create the possibility of a new or different kind of accident from any accident previously evaluated for Salem since no plant hardware modifications are required and no tests are being deleted;

c. Would not involve a significant reduction in a margin of safety, since no analytical or test changes are being made.

Conclusion The proposed technical specification changes do not involve a significant hazards consideration. The proposed changes are being implemented as a purely administrative change to achieve consistency throughout the technical specifications. This corresponds to example II.1 of guidance provided by the Commission in Federal Register FR14870 for Amendments Considered Not Likely to Involve Significant Hazards Considerations

• i DEFINITIONS

-- 1*

J thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance

.Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as

! _pump seal or valve packing leaks that are captured and conducted to

\,___,_~----a--sum--p-o-r collecting tank, or FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

SALEM - UNIT 1 1-3 Amendment No. 59

ATTACHMENT 2 TECHNICAL SPECIFICATION MARKUPS

r DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME

• 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.27 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.8 and 6.9.1.9.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

- -----~- ----.

SITE BOUNDARY C~'PS 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals,

• SALEM - UNIT 1 1-6 Amendment No. 59

I SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, F~ , of 1.55 and a reference cosine with a pea& of 1.55 for axial power shape. *n allowance is included for an increase in F H at reduced power based on the expression:

6 N

F H 6

= 1.55 [1 + 0.3(1-P)]

where P is the fraction of RATED THERMAL POWER CAPS These limiting heat flux conditions are hi er -than-those calculated for the range of all control rods full withdrawn o the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (6I) function of the Overtemperature trip. When the axial power imbalance 1

is not within the tolerance, the axial power imbalance effect on the Overtemperature 6T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping and fittings are designed to ANSI B 31.1 1955 Edition while the valves are designed to ANSI B 16.5, MSS-SP-66-1964, or ASME Section III-1968, which permit maximum transient pressures of up to 120% (2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation .

• Salem - Unit 1 B 2-2

'REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.2.2 The group demand position indicator shall be OPERABLE for each shutdown and control rod not fully inserted. IIJS6R.T 'B her-e..

APPLICABILITY: MODES 3*~ 4j~' and Sj~

ACTION:

With less than the above required group demand position indicator(s) OPERABLE, open the reactor trip system breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.2.2 Each of the above required group demand position indicator(s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days.

• With the reactor trip system breakers in the closed position

1. See S~eeial Test:- E~thm 3. 18. § 0

• SALEM - UNIT 1 3/4 1-20 Amendment No. 73

INSERT B During the performance of individual full length (shutdown and control) rod testing measurement during rod position indication system calibration:

a. Only one shutdown or control bank shall be withdrawn from the fully inserted position at a time, and

b. Keff shall be maintained less than or equal to 0.95.

1 REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from l~ s+eps tee fHlly withdrawn position shall be ~ 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T ~ 541°F, and avg

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 3.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to ~71% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

SALEM - UNIT 1 3/4 1-21

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3 .1. 3. 4 All shutdown rods shall be fully withdrawn.

---

a.I 1c~~

APPLICABILITY: MODES l'°c, and 2~'c//

ACTION:

Ca.f'!l With a maximum of one shutdown rod not fully withdraw~ except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a.

/t Ce..ps Fully withdraw the rod,(f or, y-'"

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS

c111-ps 4.1.3.4 Each shutdown rod shall be determined to be fully withdrawt;Jby use of the group demand counters, and verified by the analog rod position indicators within one hour after rod motion

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor critically, and

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

~*csee Special Test Exceptions 3 .10. 2 and 3 .10. 3

1. With Keff greater than or equal to 1.0 SALEM - UNIT 1 3/4 1-22 Amendment No. 73

(Fully Withdrawn)

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0.4 0.8 0.8 1.0 FRACTION OF RATED THERMAL.POWER Figure 3.1*1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION SALEM - UNIT 1 3/4 1-24

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• SALEM - ur: IT 1 3/~ 1-25

'3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s), and APPLICABILITY: MODE 2.

ACTION:

a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, irrunediately initiate and continue boration at ~ 10 gpm of 20,000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, irrunediately initiate and continue boration at ~ 10 gpm of 20,000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length and part length rod either partially or_!ully withdrawn,shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

~CA~

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

I SALEM - UNIT 1 3/4 10-1 Amendment No.83 I

I I

L____________ _

' REACTIVITY CONTROL SYSTEMS BASES

• 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod mis-alignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. OPERABLE condition for the analog rod position indicators is defined as being capable of indicating rod position to within +/- 12 steps of the bank demand position for a range of positions. For the Shutdown Banks, and Control Bank A this range is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 200 and 228 steps withdrawn inclusive. This permits the operator to verify that the control rods in these banks are either fully withdrawn or fully inserted, the normal operating modes for these banks. Knowledge of these banks positions in these ranges satisfies all accident analysis assumptions concerning their position. The range for control Bank B is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 160 an 228 steps withdrawn inclusive. For Control Banks C and D the range is defined as the group demand counter indicated position between 0 and 228 steps withdrawn.

Comparison of the group demand counters to the bank insertion limits with verification of rod position with the analog rod position indicators (after thermal soak after rod motion) is sufficient verification that the control rods are above the insertion limits.

The ACTION statements which perm t limited variation from the basic requirements are accompanied by addi ional restrictions which ensure that the original criteria are met. Mis-ali ment of a rod requires measurement of peaking factors or a restriction i THERMAL POWER; either of these restrictions provide assurance of uel rod integrity during continued operation. The reactivity worth f a mis-aligned rod is limited for the remainder of the fuel cycle to p event exceeding the assumption used in the accident analysis.

The maximum rod drop tim restriction is consistent with ~he assumed rod drop time used in the acciden analyses. Measurement with T >541°F and with all reactor coolant pum s operating ensures that the me~¥fired drop times will be representative of i sertion times experienced during a reactor trip at operating conditions.

Control rod positio and OPERABILITY of the rod position indicators are required to be verified n a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications equired if an automatic monitoring channel is inoperable. These ver fication frequencies are adequate for assuring that the applicable LCO's are atisfied.

______ . The full out position will be specifically established for each cycle by the Reload Safety Analysis for that cycle. This position will be within the band established by "FULLY WITHDRAWN" and will be administratively controlled. This band is allowable to minimize RCCA wear, pursuant to Information Notice 87-19. * .

SALEM UNIT 1 B 3/4 1-4 Amendemnt No. 73

' DEFINITIONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

• --ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or FULLY WITHDRAWN 1.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis *

• SALEM - UNIT 2 1-3

' DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME

• 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.27 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.8 and 6.9.1.9.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals,

• SALEM - UNIT 2 1-6

• 2. 1 SAFETY LIMITS BASES

2. 1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.*. Overheating of the fuel cladding is prevented by restricting fu~l_operation to within the nucleate boiling regime where ~he heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.* ONB is not a directly measurable parameter during operation and therafore THERMAL POWER and Reactor Coolant Temperature and Pressure.have been related to ONB thn:>ugh the W-3 correlation. The W-3 CNS correlation has been developed ta predict the DHB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, "is indicative of the margin to DNB. .

The minimum value of the DNBR during steady state operation, normal operational ~ransients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent pr.obability at a 95 percent confidence 1eve1 that CNS will not*occur and is chosen* as.an appropriate margin to CNS for a11 operating conditions.

The curves of Figures 2.1-1 and 2. 1-2 show the loci of points of THERJ'~L POWER, Reactor Coolant System pressure and average temperature for which the i minimum ONBR is no less than 1.30, or the average enthalpy at the vessel exit

, I I

is equa1 to t.he entha 1py of satYre.ted 1i quid. . - **

The curves are based on an enthalpy hot channel factor, FH , of 1.55 and a reference cosine with a peag of 1.55 for axial power shape. ~n a11owance is included for an increase in f~ at reduced power based on the expression:

F~ = 1

• SS [ 1 + O* 3 (1-P) J where P is the fraction of RATED THERMAL POWER CAPS These limiting heat flux conditi~ns are hi er than those calculated *for toe range of all control rods f w o the maximum a11owab1e control rod insertion assuming the axial power imbalance is within the limits of the f 1(de1ta I) function of the Overtemperature trip. When the axial power SALEM - UNIT 2 - B 2-l Amendment No. 20

' REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN

• LIMITING CONDITION FOR OPERATION 3.1.3.2.2 The group demand position indicator shall be OPERABLE for each shutdown and control rod not fully inserted. :C~S.e~ ~

APPLICABILITY: MODES 3*~, 4*~, and 5~

ACTION:

With less than the above required group demand position indicator(s) OPERABLE, open the reactor trip system breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.2.2 Each of the above required group demand position indicator(s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days.

• With the reactor trip system breakers in the closed position

JI.gee Special Test Exgeptia~ 3 lQ,5 SALEM - UNIT 2 3/4 1-17 Amendment No. 48

INSERT B During the performance of individual full length (shutdown and control) rod testing measurement during rod position indication system calibration:

a. Only one shutdown or control bank shall be withdrawn from the fully inserted position at a time, and

b. - Keff shall be maintained less than or equal to 0.95 *

' REACTIVITY CONTROL SYSTEMS ROD DROP TIME 2 it, LIMITING CONDITION FOR OPERATION 3:1.3.3 The individual full length (shutdown and control) rod drop time from steps the fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T greater than or equal to 541°F, and avg

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 1 & 2.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 76% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod driye system which could affect the drop time of those specific rods, and

c. At least once per 18 months.

SALEM - UNIT 2 3/4 1-18

' REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN

• LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be fully withdrawn.

APPLICABILITY: MODES l*, and 2*#

I Ct:t.pS ACTION: I With a maximum of one shutdown rod not fully/withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

cA-P.S

a. Fully/withdraw the rod,~

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS

c.Ap s 4.1.3.4 Each shutdown rod shall be determined to be fully/withdrawn by use of the group demand counters, and verified by the analog rod position indicators within one hour after rod motion

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor critically, and

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

• See Special Test Exceptions 3.10.2 and 3.10.3

1. With Keff greater than or equal to 1.0

• SALEM - UNIT 2 3/4 1-19 Amendment No. 48

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• SAt.ZM ONI'r 2 3/4 l-21 Amendment No. 34

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s).

APPLICABILITY: MODE 2.

ACTION:

a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or fully £.-A.PS_.

-CA?S withdrawn shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1 .

• SALEM - UNIT 2 3/4 10-1

__J

SPECIAL TEST EXCEPTIONS POSITION INDICATION SYSTEM - SHUTDOWN IMITING CONDITION FOR OPERATION limitations of Specification 3.1.3.2.2 may be suspende the of individual full length (shutdown and control) rod esting during rod position indication system calibration ovided:

a. shutdown or control bank the fully position at a time, and

b. Either,

1. The indicator is BLE during the withdrawal of the b.

APPLICABILITY: MODES 3, 4 performance of rod testing during rod position ACTION:

If either .10.5.a or 3. 0.5.b are not met, immediately open 4.10.5 The abov required rod position indication stems shall be determined to be OPERABLE ithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of d at least once per 24 er during rod drop time measurements by v ifying the demand ication system and the rod position indicatio agree:

12 steps when the rods are stationary, and Within 24 steps during rod motion.

SALEM - UNIT 2 3/4 10-6 Amendment No 1

' REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod mis-alignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. OPERABLE condition for the analog rod position indicators is defined as being capable of indicating rod position to within +/- 12 steps of the bank demand position for a range of positions. For the Shutdown Banks, and Control Bank A this range is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 200 and 228 steps withdrawn inclusive. This permits the operator to verify that the control rods in these banks are either fully withdrawn or fully inserted, the normal operating modes for these banks. Knowledge of these banks positions in these ranges satisfies all accident analysis assumptions concerning their position. The range for control Bank B is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 160 an 228 steps withdrawn inclusive. For Control Banks C and D the range is defined as the group demand counter indicated position between 0 and 228 steps withdrawn.

Comparison of the group demand counters to the bank insertion limits with verification of rod position with the analog rod position indicators (after thermal soak after rod motion) is sufficient verification that the control rods are above the insertion limits.b The ACTION statements which permit 1 mited variation from the basic requirements are accompanied by addftional restrictions which ensure that the original criteria are met. Mis-aligfunent of a rod requires measurement of peaking factors or a restriction i THERMAL POWER; either of these restrictions provide assurance of uel rod integrity during continued operation. The reactivity worth o a mis-aligned rod is limited for the remainder of the fuel cycle to p event exceeding the assumption used in the accident analysis.

The maximum rod drop time rest iction is consistent with the assumed rod drop time used in the accident ana ses. Measurement with T >541°F and with all reactor coolant pumps operati g ensures that the measu~¥a drop times will be representative of insertion imes experienced during a reactor trip at operating conditions.

Control rod positions and~O ERABILITY of the rod position indicators are required to be verified an a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verification. s re uired if an automatic monitoring channel is inoperable. These verific tion frequencies are adequate for assuring that the applicable LCO's are sat sfied.

I

-The full out position will be specifically established for each cycle by the Reload Safety Analysis for that cycle. This position will be within the band established by "FULLY WITHDRAWN" and will be administratively controlled. This band is allowable to minimize RC_CA_wear_, pursuant to Information Notice 87-19 .

• SALEM - UNIT 2 B 3/4 1-4 Amendment No. 48

' 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure control rod worth, and 2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the Reactor Coolant System T slightly lower than normally allowed so that the fundamental nuclear cRXfacteristics of the reactor core and related instrumentation can be verified. In order for various characteristics to be accurately measured, it is, at times, necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not be allowed by Specification 3.1.3.6 which may, in turn, cause the RCS T to fall slightly below the minimum temperature of Specification 3~Y~l.4.

3/4.10.4 NO FLOW TESTS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.

3/4.10.5 POSITION INDICATION SYSTEM-SHUTDOWN This special test exception permits the position indication systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time is derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the position indication systems remain OPERABLE.

SALEM - UNIT 2 B 3/4 10-1

ATTACHMENT 3 .

TECHNICAL SPECIFICATION CORRECTED PAGES

DEFINITIONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored.parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2 .

• FULLY WITHDRAWN l.13a FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive. FULLY WITHDRAWN will be specified in the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 1 1-3 Amendment No.

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME

• 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.27 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.8 and 6.9.1.9.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest r reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals,

• SALEM - UNIT 1 1-6 Amendment No.

SAFETY LIMITS BASES

• N The curves are based on an enthalpy hot channel factor, F H, of 1.55 and 8

a reference cosine with a pea& of 1.55 for axial power shape. Kn allowance is included for an increase in F H at reduced power based on the expression:

F~H = 1.55 [l + 0.3(1-P)]

8 where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods FULLY WITHDRAWN the maximum allowable control I

rod insertion assuming the axial power imbalance is within the limits of the f (8I) function of the Overtemperature trip. When the axial power imbalance 1

is not within the tolerance, the axial power imbalance effect on the Overtemperature 8T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping and fittings are designed to ANSI B 31.1 1955 Edition while the valves are designed to ANSI B 16.5, MSS-SP-66-1964, or ASME Section III-1968, which permit maximum transient pressures of up to 120% (2985 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation *

• Salem - Unit 1 B 2-2

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN

• LIMITING CONDITION FOR OPERATION 3.1.3.2.2 The group demand position indicator shall be OPERABLE for each shutdown and control rod not fully inserted. During the performance of individual full length (shutdown and control) rod testing measurement during rod position indication system calibration:

a. Only one shutdown or control bank shall be withdrawn from the fully inserted position at a time, and

b. Keff shall be maintained less than or equal to 0.95.

APPLICABILITY: MODES 3*, 4*, and 5*

ACTION:

With less than the above required group demand position indicator(s) OPERABLE, open the reactor trip system breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.2.2 Each of the above required group demand position indicator(s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days.

• With the reactor trip system breakers in the closed position SALEM - UNIT 1 3/4 1-20 Amendment No.

REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 228 steps withdrawn position shall be ~ 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T ~ 541°F, and avg

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 3.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to ~71% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS

• 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months .

• SALEM - UNIT 1 3/4 1-21

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN

• LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be FULLY WITHDRAWN.

APPLICABILITY: MODES 1*, and 2*#

ACTION:

With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. FULLY WITHDRAW the rod, or,

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be FULLY WITHDRAWN by use of the group demand counters, and verified by the analog rod position indicators within one hour after rod motion:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor critically, and

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

• See Special Test Exceptions 3.10.2 and 3.10.3

1. With Keff greater than or equal to 1.0 SALEM - UNIT 1 3/4 1-22 Amendment No.

• 222 -*, ...

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0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 FRACTION OF RA TED THERMAL POWER Fl~ 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER

• FOUR LOOP OPERATION

• • SALEM UNIT 1 3/4 1-24

FIGURE 3.1-2 INTENTIONALLY LEFT BLANK PENDING COMMISSION APPROVAL OF THREE LOOP OPERATION

• FIGURE 3.1-2 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOR THREE LOOP OPERATION SALEM - UNIT 1 3/4 1-25

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN

• LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s), and APPLICABILITY: MODE 2.

ACTION:

a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at ~ 10 gpm of 20,000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at ~ 10 gpm of 20,000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length and part length rod either partially or FULLY WITHDRAWN shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

SALEM - UNIT 1 3/4 10-1 Amendment No.

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod mis-alignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. OPERABLE condition for the analog rod position indicators is defined as being capable of indicating rod position to within +/- 12 steps of the bank demand position for a range of positions. For the Shutdown Banks, and Control Bank A this range is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 200 and 228 steps withdrawn inclusive. This permits the operator to verify that the control rods in these banks are either fully withdrawn or fully inserted, the normal operating modes for these banks. Knowledge of these banks positions in these ranges satisfies all accident analysis assumptions concerning their position. The range for control Bank B is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 160 an 228 steps withdrawn inclusive. For Control Banks C and D the range is defined as the group demand counter indicated position between 0 and 228 steps withdrawn.

Comparison of the group demand counters to the bank insertion limits with verification of rod position with the analog rod position indicators (after thermal soak after rod motion) is sufficient verification that the control rods are above the insertion limits. The full out position will be specifically established for each cycle by the Reload Safety Analysis for that cycle. This position will be within the band established by "FULLY WITHDRAWN" and will be administratively controlled. This band is allowable to minimize RCCA wear, pursuant to Information Notice 87-19.

The ACTION statements which permit limited variation from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. Mis-alignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restrictions provide assurance of fuel rod integrity during continued operation. The reactivity worth of a mis-aligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumption used in the accident analysis.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with T >541°F and with all reactor coolant pumps operating ensures that the me~¥fired drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified an a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied .

• SALEM - UNIT 1 B 3/4 1-4 Amendment No.

DEFINITIONS thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844 "Calculation of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for.

isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHDRAWN 1.13A FULLY WITHDRAWN shall be the condition where control and/or shutdown banks are at a position which is within the interval of 222 to 228 steps withdrawn, inclusive. FULLY WITHDRAWN will be established by the current reload analysis.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or SALEM - UNIT 2 1-3

DEFINITIONS REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE OCCURRENCE 1.27 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.9.1.8 and 6.9.1.9.

SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee, as shown in Figure 5.1-3, and which defines the exclusion area as shown in Figure 5.1-1.

SOLIDIFICATION 1.30 SOLIDIFICATION shall be the conversion of wet radioactive wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1.31 SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, SALEM - UNIT 2 1-6

2.1 SAFETY LIMITS BASES

• 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surf ace temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

N The curves are based on an enthalpy hot channel factor, F~H of 1.55 and a reference cosine with a peak ~f 1.55 for axial power shape. An allowance is included for an increase in F~H at reduced power based on the expression:

F~H = 1.55 [l + 0.3(1-P)]

where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods FULLY WITHDRAWN to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the

£ (delta I) function of the Overtemperature trip. When the axial power 1

SALEM - UNIT 2 B 2-1

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN

• LIMITING CONDITION FOR OPERATION 3.1.3.2.2 The group demand position indicator shall be OPERABLE for each shutdown and control rod not fully inserted. During the performance of individual full length (shutdown and control) rod testing measurement during rod position indication system calibration:

a. Only one shutdown or control bank shall be withdrawn from the fully inserted position at a time, and

b. Keff shall be maintained less than or equal to 0.95.

APPLICABILITY: MODES 3*, 4*, and 5*

ACTION:

With less than the above required group demand position indicator(s) OPERABLE, open the reactor trip system breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.2.2 Each of the above required group demand position indicator(s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days.

• With the reactor trip system breakers in the closed position SALEM - UNIT 2 3/4 1-17 Amendment No.

REACTIVITY CONTROL SYSTEMS ROD DROP TIME

• LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from 228 steps withdrawn shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T greater than or equal to 541°F, and avg

b. All reactor coolant pumps operating.

APPLICABILITY: MODE 1 & 2.

ACTION:

a. With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b. With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 76% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a. For all rods following each removal of the reactor vessel head,

b. For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c. At least once per 18 months .

• SALEM - UNIT 2 3/4 1-18

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN

• LIMITING CONDITION FOR OPERATION 3.i.3.4 All shutdown rods shall be FULLY WITHDRAWN.

APPLICABILITY: MODES 1*, and 2*#

ACTION:

With a maximum of one shutdown rod not FULLY WITHDRAWN, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. FULLY WITHDRAW the rod, or,

b. Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REQUIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be FULLY WITHDRAWN by use of (

the group demand counters, and verified by the analog rod position indicators within one hour after rod motion:

a. Within 15 minutes prior to withdrawal of any rods in control banks A, B, C, or D during an approach to reactor critically, and

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

• See Special Test Exceptions 3.10.2 and 3.10.3

1. With Keff greater than or equal to 1.0

• SALEM - UNIT 2 3/4 1-19 Amendment No.

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0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 FRACTION OF RA TED THERMAL POWER Ft~ J.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER FOUR LOOP OPERATION SALEM UNIT 2 3/4 1-21

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod(s).

APPLICABILITY: MODE 2.

ACTION:

a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all full length control rods inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or FULLY WITHDRAWN shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1 .

• SALEM - UNIT 2 3/4 10-1

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod mis-alignment on associated accident analyses. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. OPERABLE condition for the analog rod position indicators is defined as being capable of indicating rod position to within +/- 12 steps of the bank demand position for a range of positions. For the Shutdown Banks, and Control Bank A this range is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 200 and 228 steps withdrawn inclusive. This permits the operator to verify that the control rods in these banks are either fully withdrawn or fully inserted, the normal operating modes for these banks. Knowledge of these banks positions in these ranges satisfies all accident analysis assumptions concerning their position. The range for control Bank B is defined as the group demand counter indicated position between 0 and 30 steps withdrawn inclusive, and between 160 an 228 steps withdrawn inclusive. For Control Banks C and D the range is defined as the group demand counter indicated position between 0 and 228 steps withdrawn.

Comparison of the group demand counters to the bank insertion limits with verification of rod position with the analog rod position indicators (after thermal soak after rod motion) is sufficient verification that the control rods are above the insertion limits. The full out position will be

• specifically established for each cycle by the Reload Safety Analysis for that cycle. This position will be within the band established by "FULLY WITHDRAWN" and will be administratively controlled. This band is allowable to minimize RCCA wear, pursuant to Information Notice 87-19.

The ACTION statements which permit limited variation from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met. Mis-alignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restrictions provide assurance of fuel rod integrity during continued operation. The reactivity worth of a mis-aligned rod is limited for the remainder of the fuel cycle to prevent exceeding the assumption used in the accident analysis.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with T >541°F and with all reactor coolant pumps operating ensures that the measu~~fl drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified an a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied .

• SALEM - UNIT 2 B 3/4 1-4 Amendment No.

.3/4.10 SPECIAL TEST EXCEPTIONS BASES

• 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure control rod worth, and 2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.10.3 PHYSICS TESTS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the Reactor Coolant System T slightly lower than normally allowed so that the fundamental nuclear cRX~acteristics of the reactor core and related instrumentation can be verified. In order for various characteristics to be accurately measured, it is, at times, necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to position the various control rods at heights which may not be allowed by Specification 3.1.3.6 which may, in turn, cause the RCS T to fall slightly below the minimum temperature of Specification 3~Y~l.4.

3/4.10.4 NO FLOW TESTS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels *

• SALEM - UNIT 2 B 3/4 10-1

Attachment 4

• Background SALEM GENERATING STATIONS UNIT 1 AND 2 RCCA REPOSITIONING Public Service Electric and Gas (PSE&G) has requested a change for the control rod withdrawal limits to a range of 222 to 228 steps for*the Salem Units. The purpose of this safety evaluation is to assess the effects of this change on the LOCA related accidents addressed in the Salem FSAR. It is noted that the RCCA positioning of the fully withdrawn rods has been verified to not invalidate the current Salem control rod drop time or any other assumed tripped rod characteristics utilized in the LOCA licensing basis analyses.

BASES Small Break LQCA The small break LOCA analysis for Salem is described in FSAR Section 15.4.1 and is performed with the NRC approved Evaluation Model using the WFLASH code. This model assumes the reactor core is brought to a subcritical condition by the trip reactivity of the control rods.

Since it has been demonstrated that the rod drop time has not changed for the RCCA repositioning at the proposed rod withdrawal limits, there will be no effect on the current FSAR small break LOCA analyses

• for Salem.

Large Break LOCA The current large break LOCA analysis for Salem is described in FSAR section 15.4.1 and is performed with the NRC approved 1978 Evaluation Model and later adjusted for NUREG-0630 effects. The large break evaluation model does not take credit for the negative reactivity introduced by the control rods. During a large break LOCA, the reactor is brought to a subcritical condition by the presence of voids in the core caused by the rapid depressurization of the RCS. Since credit is not taken for the control rods, along with the fact that the rod drop time has not changed, there will be no effect on the current FSAR large break LOCA analysis for the RCCA repositioning at Salem.

Containment Integrity CLQCA Short and I.ong Term Mass and Energy Releases>

The containment mass and energy release analyses for Salem are described in FSAR Section 6.2. These sections consider the containment subcompartments and mass and energy releases for postulated LOCAs. The containment analyses, like the large break LOCA analyses, do not take credit for control rods; therefore a change to the rod withdrawal limit as proposed will have no adverse effect on the short and long term mass and energy releases presented in the Salem FSAR.

. *Steam Generator Tube Rupture The FSAR Steam Generator Tube Rupture (SGTR) analysis for Salem is presented in FSAR Section 15.4.3. The loss of coolant accident due to the primary to secondary break flow results in a decrease in the RCS pressure and reactor trip occurs on a low pressurizer pressure signal. Since the change in the rod withdrawal limits as proposed will not change the time of reactor trip and the rod drop time has been shown to not change, there will be no effect on the SGTR analysis as presented in the Salem FSAR.

Post-LQCA Long-term Core Cooling The Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 Section 50.46 Paragraph (b) Item (5) "Long Term Cooling" concludes that the reactor will remain shutdown by borated ECCS water residing in the RCS/sump after a LOCA. Since credit for the control rods is not taken for a large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a boron concentration that, when mixed with other water sources, will result in the reactor core remaining subcritical assuming all control rods out. The calculation is based upon the reactor steady state .

conditions at the initiation of a LOCA and considers sources of both borated and unborated fluid in the containment sump post-LOCA. The steady state conditions are obtained from the large break LOCA nalysis which, as stated above, do not take credit for the control rods. Thus the post-LOCA lqng-term core cooling evaluation is independent of the RCCA position as well, and therefore, there will be no change in the calculated RCS/sump Boron concentration after a postulated LOCA for Salem. The Boron related Reload Safety Analysis Checklist parameters have been verified by Westinghouse Commercial Nuclear Fuels Division Unit to remain unchanged.

Hot Lea Switchover To Prevent Potential Boron Precipitation Discussion on the hot leg switchover time is found in FSAR Sections 15.4.1 and 6.3.1.4. Post-LOCA hot leg recirculation time is determined for inclusion in emergency procedures to ensure no boron precipitation in the reactor vessel following boiling in the core.

This time is dependent on power level, and the RCS, RWST, and accumulator water volumes and boron concentrations. Since the proposed rod withdrawal limits does not affect either the power level, or the maximum boron concentrations assumed for the RCS, RWST and

• accumulator in the hot leg switchover calculation, there is no adverse effect on the post-LOCA hot leg switchover time as presented in the FSAR for Salem.

J LQCA Hydraulic Forcing Functions The LOCA hydraulic forces are considered as part of FSAR Section 3.9.3.S for Salem. Reviewing the current large break LOCA analysis reveals that.the earliest time calculated for generation of the reactor trip signal is about 0.6 seconds after break initiation.

Accounting for the unchanged rod drop time, which is on the order of several seconds, it is seen that the peak hydraulic forcing functions on the reactor vessel and interna.ls have already occurred. Typically, these peak forcing functions occur between 10 and 50 milliseconds (.01 to .05 seconds) and have subsided well before 500 milliseconds

(.SO seconds). Since the LOCA hydraulic forcing have peaked and subsided before the time the control rods are calculated to trip, the change of the control rod withdrawal limits have no effect on the LOCA hydraulic forcing functions as given in the Salem FSAR.

Conclusions

.The effect of the repos1t1oning of the fully withdrawn position of the RCCAs for the Salem Units to a range of 222 to 228 steps has been evaluated. The. potential effect on the FSAR analysis results for each of the accidents defined earlier was evaluated and shown, without exception, that the ~epC)sition~rig does no.:t _rei;;ul~ i.n cs.n_y .d.e_s.igl)_ or Regulatory lim~~-b~ing ~xc~ed~q. Therefore, it can be concluded that tbe repo~itioning of the fu1ly withdrawn position of the RCCAs for the Salem Units will be acceptable from the standpoint of the FSAR LOCA

• analyses discussed in this safety evaluation.

Since this safety evaluation supports a change to the Technical Specifications, this evaluation does not meet the provisions of 10CFRS0.59. However, this evaluation can be used to justify the implementation of this change to the technical specifica~ions.

SALEM GENERATING STATIONS UNIT l AND 2 RCCA REPOSITIONING TECHNICAL SPECIFICATION CHANGE EFFECT ON FSAR LOCA ANALYSES TABLE - l FSAR CHAPTER ACCIDENT DESCRIPTION EFFECT ON ANALYSIS 15.6.5 LARGE BREAK LOCA NO ADVERSE EFFECT ON THE FSAR PCT CALCULATIONS.

COMPLIANCE WITH 10 CFR 50.46 MAINTAINED 15.6.5 SMALL BREAK LOCA NO ADVERSE EFFECT ON THE FSAR PCT CALCULATIONS.

COMPLIANCE WITH 10 CFR 50.46 MAINTAINED 3.6 & 3.9 LOCA REACTOR VESSEL NO EFFECT ON THE LOCA AND LOOP FORCES HYDRAULIC FORCING FUNCTIONS.

POST LOCA LONG TERM CORE NO EFFECT ON THE POST-COOLING LOCA SUMP BORON CONCENTRATION.

6.3 & 15.6.5 HOT LEG SWITCHOVER TO NO EFFECT ON THE POST-PREVENT POTENTIAL BORON LOCA HOT LEG SWITCHOVER PRECIPITATION TIME.

15.6.3 STEAM GENERATOR TUBE NO EFFECT ON THE RUPTURE OFFSITE DOSES 6.2 CONTAINMENT INTEGRITY NO EFFECT ON THE (SHORT AND LONG TERM MASS MASS AND ENERGY AND ENERGY RELEASE) RELEASES

~

,;,,.: *, :-) - ATTACHMENT 4 Non-LOCA Safety Evaluation for the Salem RCCA Repositioning Background & Introduction - Public Service Electric and Gas has requested an evaluation to support changing the fully withdrawn control and shutdown rod position from 228 steps to a range of 222 to 228 steps for both Salem units. The purpose of this safety evaluation is to provide the non-LOCA justification for the requested change.

Discussion - Repositioning the fully withdrawn position from 228 steps to 222 steps has no significant impact on the Salem non-LOCA safety analysis licensing basis transients.

The RCCA repositioning could impact the non-LOCA Safety Analyses in the following areas:

Rod Drop Time - Reference 1 documents that the rod drop time (2.2 seconds to the dashpot) does not change for a fully withdrawn position of 222 to 228 steps.

RSAC Parameters - Reference 2 documents that there are no changes to the Reload Safety Analysis Checklist (RSAC).

The position of the control and shutdown banks, relative to each other in the core will not change; therefore the limiting axial power distribution assumed for the DNB analyses remain applicable. The FSAR conclusion that the DNBR design basis acceptance criteria is met for the Condition II events remains valid. Additionally, there is no significant impact on any core physics assumptions and design peaking factors important to the non-LOCA safety analyses and the reload verification.  !

I This change does not require any change to the rod control system logic.

Conclusions - It can therefore be concluded that the described chang_e do.es.

not afh.c.t. th& FSAa Hm>-LOCA_ tr_ansient behavjqr and res.ulu .as. pr.esen_t_e.d. in the FSAR and all safety analysis acceptance crtte~.ia continue. to be_ met_.

References

l. MED-RPV-1938, "Salem RCCA Axial Repositioning Drop Time Impact,*

R. R. Laubham, July 14, 1988.

2. CDF-88-127/THFL-88-333, *control Rod Repositioning for Salem Units 1 and 2,* L. Green and W. F. Weiland, July 12, 1988.