View source for RA-08-047, Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action and Clarification of a Frequency Example Using the Consolidation Line Item I

{{Adams
| number = ML081620236
| issue date = 06/09/2008
| title = Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action and Clarification of a Frequency Example Using the Consolidation Line Item Im
| author name = Cowan P
| author affiliation = AmerGen Energy Co, LLC, Exelon Generation Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000219, 05000237, 05000249, 05000254, 05000265, 05000277, 05000278, 05000373, 05000374, 05000461
| license number = DPR-016, DPR-019, DPR-025, DPR-029, DPR-030, DPR-044, DPR-056, NPF-011, NPF-018, NPF-062
| contact person = 
| case reference number = RA-08-047, RS-08-061
| document type = Letter type:RA, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Amendment
| page count = 66
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:AmerGen,.
www.exeloncorp.com Exel An Exelon Company AmerGen Energy Company, LLC Nuclear 4300 Winfield Road                                                                         Exelon Generation Warrenville, IL 60555                                                                      4300 Winfield Road Warrenville, IL 60555 10 CFR 50.90 RS-08-061 RA-08-047 June 9,2008 U. S. Nuclear Regulatory Commission AnN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50~237 and 50-249 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

==Subject:==
Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of a Frequency Example Using the Consolidated Line Item Improvement Process                            .

U.S. Nuclear Regulatory Commission June 9,2008 Page 2 of 4

==References:==
: 1. TSTF-475, Revision 1, IIControl Rod Notch Testing Frequency and SRM Insert Control Rod Action,1I dated May 22,2007
: 2. Federal Register Notice 72 FR 63935, published November 13, 2007 In accordance with the provisions of 10 CFR 50.90, IIApplication for amendment of license or construction permit,1I Exelon Generation Company, LLC, (EGC) and AmerGen Energy Company, LLC (AmerGen) are submitting a request for an amendment to the Technical Specifications (TS), Appendix A, to the Facility Operating Licenses listed above.
The proposed amendment would:
(1)  (a)    Revise the TS surveillance requirement (SR) frequency in TS 3.1.3, "Control Rod OPERABILITY" (except for Oyster Creek Nuclear Generating Station).
(b)    Revise the TS surveillance requirement in TS 4.2, IIReactivity Control,1I Specification 0 (for Oyster Creek Nuclear Generating Station).
(2)  Clarify the reqUirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, Required Action E.2, "Source Range Monitoring Instrumentation" (Clinton Power Station only).
(3)  Revise Example 1.4-3 in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension (Oyster Creek Nuclear Generating Station excluded). provides a description of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides the existing TS Bases pages marked up to reflect the proposed changes (for information only). Attachment 4 provides a summary of the regulatory commitments made in this submittal.
The proposed changes have been reviewed by the Plant Operations Review Committee at each of the stations and approved by their respective Nuclear Safety Review Boards in accordance with the requirements of the EGC and AmerGen Quality Assurance Programs.
EGC and AmerGen request approval of the proposed License Amendments by June 9, 2009, with the amendments being implemented within 60 days of issuance.
In accordance with 10 CFR 50.91 (b)(1), IINotice for public comment; State consultation, II EGC and AmerGen are notifying the States of Illinois and New Jersey, and the Commonwealth of Pennsylvania of this application for changes to the TSs by transmitting a copy of this letter and its attachments to the designated State Officials.

U.S. Nuclear Regulatory Commission June 9, 2008 Page 3 of 4 Should you have any questions concerning this letter, please contact Mr. Frank Mascitelli at (610) 765-5512.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 9th day of June 2008.
Respectfully, Cff
    !1r~&L~
Pamela B. Cowan Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC AmerGen Energy Company, LLC Attachments: 1.      Description and Assessment
: 2. Proposed Technical Specification Changes
: 3. Proposed Technical Specification Bases Changes (For Information Only)
: 4. List of Regulatory Commitments cc:   Regional Administrator, Region I, USNRC Regional Administrator, Region III, USNRC NRC Project Manager, NRR - Clinton Power Station NRC Project Manager, NRR - Dresden Nuclear Power Station NRC Project Manager, NRR - LaSalle County Station NRC Project Manager, NRR - Oyster Creek Nuclear Generating Station NRC Project Manager, NRR - Peach Bottom Atomic Power Station NRC Project Manager, NRR - Quad Cities Nuclear Power Station Senior Resident Inspector - Clinton Power Station Senior Resident Inspector - Dresden Nuclear Power Station Senior Resident Inspector - LaSalle County Station Senior Resident Inspector - Oyster Creek Nuclear Generating Station Senior Resident Inspector - Peach Bottom Atomic Generating Station Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Resources Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection Mayor of Lacey Township, Forked River, NJ S. T. Gray, State of Maryland

ATTACHMENT 1 Description and Assessment Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of a Frequency Example Using the Consolidated Line Item Improvement Process

==1.0    DESCRIPTION==

2.0    ASSESSMENT 2.1    Applicability of Published Safety Evaluation 2.2    Optional Changes and Variations

==3.0    REGULATORY ANALYSIS==

3.1    No Significant Hazards Consideration Determination 3.2    Verification and Commitments 4.0    ENVIRONMENTAL EVALUATION

==5.0    REFERENCES==

Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and                                        ATTACHMENT 1 Clarification of a Frequency Example Using the Consolidated                  Description and Assessment Line Item Improvement Process                                                                      Page 2

==1.0    DESCRIPTION==

In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license or construction permit,'* Exelon Generation Company, LLC, (EGC) and AmerGen Energy Company, LLC (AmerGen) hereby request the following amendment to the Technical Specifications (TS), Appendix A, for the following Facility Operating Licenses:
Dresden Nuclear Power Station, Units 2 and 3               Renewed Facility Operating License Nos. DPR-19 and DPR-25 LaSalle County Station, Units 1 and 2                     Facility Operating License Nos. NPF-11 and NPF-18 Peach Bottom Atomic Power Station, Units 2                Renewed Facility Operating License and 3                                                     Nos. DPR-44 and DPR-56 Quad Cities Nuclear Power Station, Units 1                Renewed Facility Operating License and 2                                                     Nos. DPR-29 and DPR-30 AmerGen Clinton Power Station, Unit 1                             Facility Operating License No. NPF-62 Oyster Creek Nuclear Generating Station                   Facility Operating License No. DPR-16 The proposed amendment would:
(1) (a) Revise the TS surveillance requirement (SR) frequency in TS 3.1.3, "Control Rod OPERABILITY" (except for Oyster Creek Nuclear Generating Station).
(b) Revise the TS surveillance requirement in TS 4.2, "Reactivity Control," Specification D (for Oyster Creek Nuclear Generating Station).
(2) Clarify the requirement to fully insert all insertable control rods for the limiting condition for operation (LCO) in TS 3.3.1.2, Required Action E.2, "Source Range Monitoring Instrumentation" (Clinton Power Station only).
(3) Revise Example 1.4-3 in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension (Oyster Creek Nuclear Generating Station excluded).
The changes are consistent with Nuclear Regulatory Commission (NRC) approved Industryffechnical Specification Task Force (TSTF) Standard Technical Specification (STS) change TSTF- 475, Revision 1 (Reference 5.1). The Federal Register Notice (Reference 5.2) published on November 13, 2007 announced the availability of these TS improvements through the consolidated line item improvement process (CLlIP). The proposed changes are consistent with Reference 5.1 , with the only exceptions being specified in Section 2.2.

Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and                                     ATTACHMENT 1 Clarification of a Frequency Example Using the Consolidated                 Description and Assessment Line Item Improvement Process                                                                   Page 3 2.0    ASSESSMENT 2.1       Applicability of Published Safety Evaluation EGC and AmerGen have reviewed the safety evaluation dated November 13,2007, as part of the CUIP. This review included a review of the NRC staff*s evaluation, as well as the supporting information provided to support TSTF-475, Revision 1. EGC and AmerGen have concluded that the justifications presented in the TSTF proposal and the safety evaluation prepared by the NRC staff are applicable to Clinton Power Station, Unit 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 and justify this amendment for the incorporation of the changes to the aforementioned stations* TS.
2.2       Optional Changes and Variations EGC and AmerGen are not proposing any significant variations or deviations from the TS changes described in the TSTF-475, Revision 1 and NRC staffs model safety evaluation dated November 13, 2007.
Four minor variations are described as follows:
Oyster Creek Nuclear Generating Station is a custom technical specification BWR/2 plant and, therefore, the applicable TSs and associated bases sections and wording are different from the BWRl4 and BWRl6 STSs. The minor variations are grammatical and administrative in nature and do not change the technical intent of the changes proposed for control rod operability surveillance frequency requirements for SR 3.1.3.2. The proposed SR frequency does not deviate from the TSTF-proposed 31-day frequency. In addition, since the Oyster Creek Nuclear Generating Station technical specifications are not STSs, the SRM Insert Control Rod Action clarification and the Example 1.4-3 Section 1.4, IIFrequencyll clarification are not required for Oyster Creek Nuclear Generating Station.
Due to the large number of procedure changes that would be required to be revised due to the renumbering of the subsequent SRs following SR 3.1.3.2 deletion, SR 3.1.3.2 will be reflected as deleted and the subsequent SRs will not be renumbered as described in the approved TSTF-475, Revision 1. As a result, the associated administrative reference changes due solely to the renumbered SRs are not required.
For Clinton Power Station, the NOTE associated with SR 3.1.3.3 is being revised to be consistent with the STS and the changes associated with Example 1.4-3 included in this amendment request.
The proposed amendment does not adopt the clarification of Source Range Monitor (SRM) TS action for inserting control rods for Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2. These station's TS and

Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and                                           ATTACHMENT 1 Clarification of a Frequency Example Using the Consolidated                       Description and Assessment Line Item Improvement Process                                                                          Page 4 associated TS Bases concerning TS Section 3.3.1.2 required Action E.2 currently have the clarification to fully insert all insertable control rods for the limiting condition for operation.

==3.0       REGULATORY ANALYSIS==

3.1       No Significant Hazards Consideration Determination EGC and AmerGen have reviewed the proposed no significant hazards consideration determination (NSHCD) published in the Federal Register as part of the CUIP. EGC and AmerGen have concluded that the proposed NSHCD presented in the Federal Register Notice on November 13, 2007, is applicable to Clinton Power Station, Unit 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 and is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).
3.2       Verification and Commitments As discussed in the notice of availability published in the Federal Register on November 13, 2007, for this TS improvement, EGC and AmerGen verified the applicability of TSTF-475, Revision 1 to Clinton Power Station, Unit 1, Dresden Nuclear Power Station, Units 2 and 3, LaSalle County Station, Units 1 and 2, Oyster Creek Nuclear Generating Station, Peach Bottom Atomic Power Station, Units 2 and 3, and Quad Cities Nuclear Power Station, Units 1 and 2 and commits to establishing Technical Specification Bases for TS as proposed in TSTF-475, Revision 1.
These changes are based on TSTF change traveler TSTF-475 (Revision 1), that proposes revisions to the STS by: (1) Revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod, from 117 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM [118 days 18 hours after the control rod is fully withdrawn and THERMAL POWER is greater than the LPSP of RPCSII (Clinton Power Station) and IIEach partially or fully withdrawn control rod shall be exercised at least once each week ll (Oyster Creek Nuclear Generating Station)f' to 1131 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM [1131 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RPCSII (Clinton Power Station) and IIEach withdrawn control rod shall be exercised at least once each month ll (Oyster Creek Nuclear Generating Station)], II (2) adding the word IIfullyll to LCO 3.3.1.2 Required Action E.2 to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated SRM instrument is inoperable (Clinton Power Station only), and (3) revising Example 1.4-3 in Section 1.4 IIFrequencyll to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the IISURVEILLANCE II column in addition to the time periods in the IIFREQUENCYII column (Oyster Creek Nuclear Generating Station excluded).

Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and                                  ATTACHMENT 1 Clarification of a Frequency Example Using the Consolidated              Description and Assessment Line Item Improvement Process                                                                 PageS 4.0      ENVIRONMENTAL EVALUATION EGC and AmerGen have reviewed the environmental evaluation included in the model safety evaluation dated November 13, 2007, as part of the CUIP. EGC and AmerGen have concluded that the NRC staff's findings presented in that evaluation are applicable to Clinton Power Station, Unit 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 and the evaluation is hereby incorporated by reference for this application.

==5.0      REFERENCES==

5.1      TSTF-475, "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action,"
Revision 1.
5.2      "Notice of Availability of Model Application Concerning Technical Specification Improvement To Revise Control Rod Notch Surveillance Frequency, Clarify SRM Insert Control Rod Action, and Clarify Frequency Example, published in Federal Register/ Vol.
II 72, No. 218, November 13, 2007.

ATTACHMENT 2 Proposed Technical Specification Changes

TECHNICAL SPECIFICATION PAGES (Mark-ups)
Clinton Power Station. Unit 1 Tech nical Specification Pages 1.0-27 1.0-28 3.1-7 3.1-9 3.1-10 3.3-11

Frequency 1.4 1.4  Frequency EXAMPLES       EXAMPLE 1.4-2     (continued)
                 "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                    FREQUENCY
                  ------------------NOTE------------------
Not required to be performed until 12 hours after ~ 25% RTP.
Perform channel adjustment.                7 days The interval continues whether or not the unit operation is
                < 25% RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power ~ 25% RTP.
(continued)
( PLVS-rHc;   EXTb7VSJD,v .ALLOvJ&-o (Jy 5(2. 3. C>. ~)
CLINTON                             1.0-27                    Amendment No. 95

1.4  Frequency EXAMPLES        EXAMPLE 1.4-3    (continued)
Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE                   FREQUENCY
                  ------------------NOTE------------------
Only required to be met in MODE 1.
Verify leakage rates are within limits. 24 hours Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.
Therefore, if the Surveillance were not performed within the 24 hour (plus the extension allowed by SR 3.0.2) interval, but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
CLINTON                             1.0-28                  Amendment No. 95

Control Rod OPERABILITY 3.1.3 3.1   REACTIVITY CONTROL SYSTEMS 3.1.3    Control Rod OPERABILITY LCO   3.1.3        Each control rod shall be OPERABLE.
APPLICABILITY:     MODES 1 and 2.
ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each control rod.
CONDITION                    REQUIRED ACTION              COMPLETION TIME A. One withdrawn control     ------------NOTE-------------
rod stuck.                A stuck rod may be bypassed in the Rod Action Control System (RACS) in accordance with SR 3.3.2.1.9 if required to allow continued operation.
A.l      Disarm the associated       2 hours control rod drive (CRD) .
AND A.2      Perform (SR 3/1. 3 . D      24 hours from
                                      ~SR 3.1.3.3 for              discovery of each withdrawn              Condition A OPERABLE control rod.       concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the Rod Pattern Control System (RPCS)
AND A.3      Perform SR 3.1.1.1.         72 hours (continued)
CLINTON                                 3.1-7                        Amendment No. 95

Control Rod OPERABILITY 3.1.3 ACTIONS (Continued)
CONDITION                   REQUIRED ACTION           COMPLETION TIME E. Required Action and      E.1      Be in MODE 3.           12 hours associated Completion Time of Condition A, C, or D not met.
OR Nine or more control rods inoperable.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                              FREQUENCY SR   3.1.3.1     Determine the position of each control rod. 24 hours
                 ----------- -------NOTE-------
Not require to be performed      til 8 days 18 hours a ter the control ro is fully withdrawn nd THERMAL POWER s greater than the LPSP f the RPCS.
7 days (continued)
CLINTON                                3.1-9                Amendment No. 95T 102

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE                             FREQUENCY SR   3.1.3.3    -------------------NOTE----------------
Not re-.Sluired to be performed until @ days       3, r18jhoursVafter the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RPCS.
Insert each (pariialjjO withdrawn control rod    31 days at least one no ch.
SR   3.1.3.4   Verify each control rod scram time from           In accordance fully withdrawn to notch position 13 is          with
:::;; 7 seconds.                                  SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 SR   3.1.3.5   Verify each control rod does not go to the        Each time the withdrawn overtravel position.                    control rod is withdrawn to "full out" position AND Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling CLINTON                                 3.1-10              Amendment No. 95, 102

SRM Instrumentation 3.3.1.2 ACTIONS   (continued)
CONDITION                REQUIRED ACTION         COMPLETION TIME D. One or more required D.1    Fully insert all         1 hour SRMs inoperable in          insertable control MODE 3 or 4.                rods.
AND D.2    Place reactor mode       1 hour switch in the shutdown position.
E. One or more required E.1    Suspend CORE             Immediately SRMs inoperable in          ALTERATIONS except MODE 5.                      for control rod insertion.
AND
                           --                       ~)
E.2     Initiate action  to~    Immediately insert all insertable control rods in core cells containing one or more fuel assemblies.
CLINTON                        3.3-11                      Amendment No. 95

TECHNICAL SPECIFICATION PAGES (Mark-ups)
Dresden Nuclear Power Station. Units 2 and 3 Technical Specification Pages 1.4-4 3.1.3-2 3.1.3-4

Frequency 1.4 1.4   Frequency EXAMPLES           EXAMPLE 1.4-3 (continued)
SURVEILLANCE REQUIREMENTS SURVEILLANCE                    FREQUENCY
                     ------------------NOTE-----------------
Not required to be performed until 12 hours after ~ 25% RTP.
Perform channel adjustment.                 7 days The interval continues whether or not the unit operation is
                  < 25%   RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Shoul d the 7 day interval be exceeded whi 1e operation is < 25% RTP, this Note allows 12 hours after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power ~ 25% RTP.
Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed wi thi n thi s 12 hour i nterv~, there woul d then be a failure to perform a urvel ~- wlthin the specified Frequency, and the provisions of SR 3.0.3 would apply.
(continued)
(PLUS THE G'X"TEN510N (3y    SIC.. 6~O.~)
Dresden 2 and 3                        1. 4-4               Amendment No. 185/180

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION             REQUIRED ACTION           COMPLETION TIME A.   (cont inued)         A.3   Perform<&sect;R 3/1.3:])     24 hours from
                                 ~ SR 3.1.3.3 for        discovery of each withdrawn          Condition A OPERABLE control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM A.4   Perform SR 3.1.1.1. 72 hours B. Two or more withdrawn B.1   Be in MODE 3.           12 hours control rods stuck.
C. One or more control   C.1   --------NOTE---------
rods inoperable for         RWM may be bypassed reasons other than          as allowed by Condition A or B.           LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.
Fully insert            3 hours inoperable control rod.
C.2   Disarm the associated   4 hours CRD.
(continued)
Dresden 2 and 3                3.1.3-2               Amendment No. 185/180

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                         FREQUENCY SR  3.1.3.1     Determine the position of each control rod.                                        24 hours SR               - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - -
Not requir d to be perfo ed until 7 days after the control rod is withdrawn and THERMAL P WER is greate than the LPSP of RWM.
SR  3.1.3.3      - - - - - - - - - - - - - - - - - - -NOTE - - - - - - - - - - - - - - - - - - - -
Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
Insert each r~aii.0withdrawn control rod                                          31 days at least one no en.
SR  3.1.3.4    Verify each control rod scram time from                                             In accordance fully withdrawn to 90% insertion is                                                with
: : ; 7 seconds.                                                                    SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
Dresden 2 and 3                                         3.1.3-4                                Amendment No. 185/180

TECHNICAL SPECIFICATION PAGES (Mark-ups)
LaSalle County Station. Units 1 and 2 Technical Specification Pages 1.4-4 3.1.3-2 3.1.3-4

Frequency 1.4 1.4   Frequency EXAMPLES           EXAMPLE 1.4-3 (continued)
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                       FREQUENCY
                     - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - --
Not required to be performed until 12 hours after ~ 25% RTP.
Perform channel adjustment.                                                7 days The interval continues whether or not the unit operation is
                   <  25% RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Shoul d the 7 day interval be exceeded whi 1e operation is < 25% RTP, this Note allows 12 hours after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided
__---  o,peration does not exceed 12 hours with power ~ 25% RTP.
would be allowed for If the Surveillance were not interva , there would then be
                --~-T~~~~~~~~--~~~~a~n~c~e~wTithin                                              the specified of SR 3.0.3 would apply.
(continued)
LaSalle 1 and 2                                        1.4-4                               Amendment No. 147/133

Control Rod OPERABI LITY 3.1. 3 ACTIONS CONDITION              REQUIRED ACTION               COMPLETION TIME A.  (continued)           A.3    PerformLSR   3/1. 3])       24 hours from
                                 ~SR 3.1.3.3 for              discovery of each withdrawn              Condition A OP ERAB LE cont ro 1 rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM A.4    Perform SR 3.1.1.1.         72 hours B. Two or more withdrawn B.1    Be in MODE 3.               12 hours control rods stuck.
C. One or more control   C.l    --------NOTE---------
rods inoperable for          RWM may be bypassed reasons other than           as allowed by Condition A or B.            LCO 3.3.2. 1, if required, to allow insertion of inoperable control rod and continued operation.
Fully insert                3 hours inoperable control rod.
C.2    Disarm the associated       4 hours CRD.
(continued)
LaSalle 1 and 2                3.1.3-2                   Amendment No. 147/133

Control Rod OPERABILITY 3.1. 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                         FREQUENCY SR  3.1.3.1      Determine the position of each control rod.                                      24 hours SR  3.1.3.2      - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - - - -
Not required 0 be performed ntil 7 days after the co trol rod is with rawn and THERMAL POW R is greater tha the LPSP of the RWM.
SR  3.1.3.3      -------------------NOTE--------------------
Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
Insert each [partfallJUwithdrawn control rod                                       31 days at least one notch.
SR  3.1.3.4     Verify each control rod scram time from                                           In accordance fully withdrawn to notch position 05 is                                           with
::s; 7 seconds.                                                                    SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
LaSalle 1 and 2                                       3.1.3-4                                 Amendment No. 147/133

TECHNICAL SPECIFICATION PAGES (Mark-ups)
Oyster Creek Nuclear Generating Station Technical Specification Page 4.2-1

4.2    REACTIVITY CONTROL Applicability:    Applies to the surveiBance requirements for reactivity control.
Objective:        To verify the capability for controlling reactivity.
Specification:
A.       SOM shall be verified:
: 1.       Prior to each CORE ALTERATION, and
: 2.       Once within 4 hours following the first criticality following any CORE ALTERATION.
B.       The control rod drive housing support system shall be inspected after reassembly.
C.       The maximum scram insertion time of the control rods shall be demonstrated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:
: 1.        For all control rods prior to thermal power exceeding 40% power with reactor coolant pressure greater than 800 psig, following core alteration,s or after a reactor shutdown that is greater than 120 days.
: 2.       For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "a" or "b" as follows:
a.1      Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90 % insertion shall not exceed 2.2 seconds, and a.2       Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 40% power.
: b.        Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure.
: 3.       On a frequency of less than or equal to once per 180 days of cumulative power operation, for at least 20 control rods, on a rotating basis, with reactor coolant pressure greater than 800 psig.
D.       Each parti lIy 0 full withdrawn control rod shall be exercised c* least once each w ek.        IS tes shall be performed within 24 hours in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one 1Jlly or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out. The surveillance need not be completed within 24 hours if the number of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.
OYSTER CREEK                                     4.2-1                   Amendment No: 178, 198, 249

TECHNICAL SPECIFICATION PAGES (Mark-ups)
Peach Bottom Atomic Power Station, Unit 2 Technical Specification Pages 1.4-4 1.4-5 3.1-8 3.1-10

Frequency 1.4 1.4 Frequency EXAMPLES       EXAMPLE 1.4-2 (continued)
                "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25S RTP.
EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                  FREQUENCY
                 ------------------NOTE------------------
Not required to be performed until 12 hours after ~ 25% RTP.
Perform channel adjustment.               7 days The interval continues whether or not the unit operation is
               < 25S RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 251 RTP, this Note allows 12 hours after power reaches ~ 251 RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LeO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours~with power ~ 25% RTP.
( PLUS il-t&#xa3; e'XTEIJSlOIU G'i   SR 3~o~~)
PBAPS UNIT 2                        1.4-4                  Amendment No. 210

Frequency 1.4 1.4 Frequency EXAMPLES       EXAMPLE 1.4-3 (continued)
Once the unit reaches 25~ RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour interval, ere wou        en e a failure to perform a Surveillance within the specified Frequency, and th~ p~ovisions of SR 3.0.3 would apply.
(PLUS ~ I:XT':-NS/OA../ ALLOVJ(;-1)
EXAMPLE 1.4-4       8y Srz.. 3,O,J...)
SURVEILLANCE REQUIREMENTS -------------
SURVEILLANCE                   FREQUENCY
                 ------------------NOTE------------------
Only required to be met in MODE 1.
Verify leakage rates are within limits. 24 hours Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance.
Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2),
but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCD. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
PBAPS UNIT 2                       1.4-5                   Amendment No. 210

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION            REQUIRED ACTION            COMPLETION TIME A.   (continued)          A.3   Perform ~ 3J .3
                                 ~ SR
::v 3.1.3.3 for 24 hours from discovery of each withdrawn           Condition A OPERABLE control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND A.4    Perform SR 3.1.1.1.      72 hours B. Two or more withdrawn B.l   Be in MODE 3.             12 hours control rods stuck.
C. One or more control   C.l    --------NOTE---------
rods inoperable for        RWM may be bypassed reasons other than         as allowed by Condition A or B.          LCO 3.3. 2.1, if required, to allow insertion of inoperable control rod and continued operation.
              -                  ~--~~----------------
Fully insert              3 hours inoperable control rod.
AND C.2   Disarm the associated     4 hours CRD.
(continued)
PBAPS UNIT 2                    3.1-8                      Amendment No. 210

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY SR 3.1.3.1     Determine the position of each control rod.        24 hours SR              ------------ ------NOTE--------------------
Not require to be perfo d until 7 days after the c ntrol rod is ithdrawn and THERMAL PO ER is greater han the LPSP of the RWM.
Insert e ch fully with rawn control least 0 notch.
SR 3.1.3.3      -------------------NOTE--------------------
Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
Insert each ce!!-~ alii> wi thdrawn control rod    31 days at least one no cfi.
SR 3.1.3.4     Verify each control rod scram time from             In accordance fully withdrawn to notch position 06 is             with
::s 7 seconds.                                       SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
PBAPS UNIT 2                            3.1-10                     Amendment No. 210

TECHNICAL SPECIFICATION PAGES (Mark-ups)
Peach Bottom Atomic Power Station, Unit 3 Technical Specification Pages 1.4-4 1.4-5 3.1-8 3.1-10

Frequency 1.4 1.4 Frequency EXAMPLES       EXAMPLE 1.4-2 (continued)
                "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                  FREQUENCY
                 ------------------NOTE------------------
Not required to be performed until 12 hours after ~ 25% RTP.
Perform channel adjustment.               7 days The interval continues whether or not the unit operation is
               < 25% RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0~4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power ~ 25% RTP.
(PLUS        THE EXTENSIDf'J f2:>'i <;(1. 3.0 ~ ~ )
PBAPS UNIT 3                          1.4-4                Amendment No. 214

Frequency 1.4 1.4 Frequency EXAMPLES       EXAMPLE 1.4-3 (continued)
Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not.
performed within this 12 hour interva, ere wou             en be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.
( PLU 5 TH-tZ t='XTg..JS; () tJ  ALLO ()J(;:-""D .
EXAMPLE 1.4-4        r?>Y srt 3.0" :t.)
SURVEILLANCE REQUIREMENTS SURVEILLANCE                       FREQUENCY
                 ------------------NOTE------------------
Only required to be met in MODE 1.
Verify leakage rates are within limits.      24 hours Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an Rotherwise stated R exception to the Applicability of this Surveillance.
Therefore, if the Surveillance were not performed within the 24 hour interval (plus the extension allowed by SR 3.0.2),
but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour Frequency were not met), SR 3.0.4 would require satisfying the SR.
PBAPS UNIT 3                        1.4-5                    Amendment No. 214

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION             REQUIRED ACTION            COMPLETION TIME A.   (continued)           A.3  ~rm~ 3/1.3.2              24 hours from a     R 3.1.3.3 for      discovery of e ch withdrawn           Condition A OPERABLE control rod. concurrent with THERMAL POWER greater than the low power setpoint (lPSP) of the RWM AND A.4    Perform SR 3.1.1.1.      72 hours B. Two or more withdrawn B.1    Be in MODE 3.            12 hours control rods stuck.
C. One or more control   C.1    --------NOTE---------
rods inoperable for         RWM may be bypassed reasons other than          as allowed by Condition A or B.           LCO 3.3.2. 1, i f required, to allow insertion of inoperable control rod and continued operation.
Fully insert              3 hours inoperable control rod.
AND C.2   Disarm the associated     4 hours CRD.
(continued)
PBAPS UNIT 3                     3.1-8                     Amendment No. 214

Control Rod OPERABILITY 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                             FREQUENCY SR 3.1.3.1    Determine the position of each control rod. 24 hours
                  ------------- -----NOTE----- --------------
Not required 0 be performed until 7 days after the co trol rod is wi drawn and THERMAL POW is greater th n the LPSP of the RWM.
Insert ea h fully withdra n control             7 days least on notch.
SR 3.1.3.3     ----------------~--NOTE--------------------
Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
Insert each~r~riil)withdrawn control rod        31 days at least one no ch.
SR 3.1.3.4     Verify each control rod scram time from         In accordance fully withdrawn to notch position 06 is         with
:s 7 seconds.                                    SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)
PBAPS UNIT 3                          3.1-10                   Amendment No. 214

TECHNICAL SPECIFICATION PAGES (Mark-ups)
Quad Cities Nuclear Power Station. Units 1 and 2 Technical Specification Pages 1.4-4 3.1.3-2 3.1.3-4

Frequency 1.4 1.4 Frequency EXAMPLES           EXAMPLE 1.4-3 (continued)
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY
                    - - - - - - - - - - - - - - - - - - NOT E- - - - - - - - - - - - - - - - -
Not required to be performed until 12 hours after ~ 25% RTP.
Perform channel adjustment.                                                 7 days The interval continues whether or not the unit operation is
                  <   25% RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours with power ~ 25% RTP.
Once the unit reaches 25% RTP, 12 hours would be allowed for completing the Surveillance. If the Surveillance were not er               within this 12 hour interval, there would then be a failure to perform a urvel                                  ce within the specified Frequency, and the provisions of SR 3.0.3 would apply.
(continued)
               '(PLUS               -r~ ~XT\:7USlarJ elY        S1t 3!o.~)
Quad Cities 1 and 2                                   1. 4-4                                Amendment No. 199/195

Control Rod OPERABILITY 3.1.3 ACTIONS CONDITION             REQUIRED ACTION           COMPLETION TIME A.   (continued)          A.3   PerformcfR  3h.3.D      24 hours from
                                 ~SR 3.1.3.3 for         discovery of each withdrawn          Condition A OPERABLE control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM A.4   Perform SR 3.1.1.1. 72 hours B. Two or more withdrawn B.l   Be in MODE 3.           12 hours control rods stuck.
C. One or more control   C.l   --------NOTE---------
rods inoperable for         RWM may be bypassed reasons other than          as allowed by Condition A or B.           LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.
Fully insert            3 hours inoperable control rod.
C.2   Disarm the associated   4 hours CRD.
(continued)
Quad Cities 1 and 2            3.1.3-2               Amendment No. 199/195

Control Rod OPERABILITY 3.1. 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY SR  3.1.3.1     Determine the position of each control rod.                                    24 hours SR              -------------------NOTE--------- ----------
Not required to be performed until 7 days after the co trol rod is wit drawn and THERMAL POWE is greater th                              the LPSP of RWM.
withdrawn control rod SR  3.1.3.3     - - - - - - - - - - - - - - - - - - - NOTE - - - - - - - - - - - - - - - - - -
Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.
Insert eachee;rtia(G)withdrawn control rod                                     31 days at least one notch.
SR  3.1.3.4    Verify each control rod scram time from                                         In accordance fully withdrawn to 90% insertion is                                             with
                ~ 7 seconds.                                                                    SR 3.1.4.1.
SR 3.1. 4.2.
SR 3.1.4.3. and SR 3.1.4.4 (continued)
Quad Cities 1 and 2                                    3.1.3-4                             Amendment No. 199/195

ATTACHMENT 3 Proposed Technical Specification Bases Changes (For Information Only)

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
Clinton Power Station, Unit 1 Technical Specification Bases Pages B 3.1-16 B 3.1-19 B 3.1-20 B 3.3-35

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS  A.1, A.2, and A.3        (continued) manner. Isolating the control rod from scram prevents damage to the CRDM.        The control rod can be isolated from scram by isolating the hydraulic control unit from scram and normal drive and withdraw pressure, yet still maintain cooling water to the CRD.
Monitoring of the insertion capability for each withdrawn control rod must also be performed_within 24 hours.                 ~
(SR  3.X. 3::o:JilanwSR 3.1.3.3 perform"""periocHc tests of the       '2/
control rod insertion capability of withdrawn control rods.
Testing each withdrawn control rod ensures that a generic problem does not exist. The allowed Completion Time of 24 hours provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.       Required Action A.2 has a modified time zero Completion Time.       The 24 hour Completion Time for this Required Action starts when the withdrawn control rod is discovered to be stuck and THERMAL POWER is greater than the actual low power setpoint (LPSP) of the rod pattern controller (RPC) , since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RPC (LCO 3.3.2.1, "Control Rod Block Instrumentation") .
To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours.      Should a DBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required.      Therefore, the original SDM demonstration may not be valid.      The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.
The allowed Completion Time of 72 hours to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity.       Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram.         Even with the postulated additional single failure of an adjacent control (continued)
CLINTON                        B 3.1-16                     Revision No. 1-1

Control Rod OPERABILITY B 3.1.3 BASES   (continued)
SURVEILLANCE        SR  3.1.3.1 REQUIREMENTS The position of each control rod must be determined, to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.
(?R   3 .. L3; ~         DELE"r<~
c(( 13 .1 . 3Z2 ai9> SR       3.1.3.3
(&sect;)
         @-__--r-t::""'::~:::::-::-~=-=-=-=-:r:
CLINTON                                     B 3.1-19             Revision No. 2-14

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE REQUIREMENTS (ff 73 .1.-3/2 ~ SR   3.1.3.3   (continued) immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.
SR  3.1.3.4 Verifying the scram time for each control rod to notch position 13 is ~ 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown functions.
This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV)
Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
With regard to scram time values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 9).
SR  3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The Surveillance requires verifying that a control rod does not go to the withdrawn overtravel position when it is fUlly withdrawn. The overtravel position feature provides a positive check on the coupling integrity, since only an uncoupled CRD can reach the overtravel position. If the control rod goes to the withdrawn overtravel position, the control rod drive mechanism can be inserted to attempt recoupling, within the limitations of Condition C. This verification is required (continued)
CLINTON                        B 3.1-20                     Revision No. 4 6

SRM Instrumentation B 3.3.1.2 BASES ACTIONS        E.1 and E.2 (continued) with one or more required SRMs inoperable in MODE 5, the capability to detect local reactivity changes in the core during refueling is degraded. CORE ALTERATIONS must be immediately suspended, and action must be immediately initiated t~insert all insertable control rods in core cells containing one or more fuel assemblies. Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity, given that fuel is present in the core. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.
Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.
SURVEILLANCE   The SRs for each SRM Applicable MODE or other specified REQUIREMENTS   condition are found in the SRs column of Table 3.3.1.2-1.
SR   3.3.1.2.1 and SR   3.3.1.2.3 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to the same parameter indicated on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
(continued)
CLINTON                       B 3.3-35                        Revision No. a

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
Dresden Nuclear Power Station. Units 2 and 3 Technical Specification Bases Pages B 3.1.3-4 B 3.1.3-8

Control Rod OPERABI LITY B 3.1.3 BASES ACTIONS          A.l. A.2. A.3. and A.4   (continued)
RWM is bypassed to ensure compliance with the CRDA analysis.
With one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The stuck control rod separation criteria are not met if: a) the stuck control rod occupies a location adjacent to two "slow" control rods, b) the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pai r of "slow" control rods e1 sewhere in the core adjacent to one another. The description of "slow" control rods is provided in LCO 3.1.4 "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours. The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CRDM or reactor internals. The control rod isolation method should also ensure cooling water to the CRD is maintained.
Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.       ~
QR 3.143.2 ji@)SR 3.1. 3.3 perform<Peri odi c tests of the contra rod insertion capability of withdrawn control rods.
Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and (continued)
Dresden 2 and 3                   B 3.1.3-4                          Revision  a

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE    @    f.1.3.yan&sect;J SR      3.1.3.3 REQUIREMENTS (continued) Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
The control rod may then be returned to its original position. This ensures the control rod is not stuck and free to insert on a scram si gna1 .       h s Survei 11 ance not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the analyzed rod osition se uence (LCO 3.1.6) and the RWM (LCO 3.3.2.1).
The 7 day Fre uency of SR 3. . . 2 is based on oper ting experience r 1ated to the cha ges in CRD performa ce and the ease of perf rming notch tes 'ng for fully withd awn control rods. Part ally withdrawn c ntro1 rods are tes ed at a 31 day Fre uency, based on         e potential power eduction required t allow the contr 1 rod movement and considering the large testing sample 0 SR 3.1.3.2. Furt ermore, th~
      ~ . ~31 daiFrequency takes into account operating experience
      ~ . related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and
: a. pp.ropriate acti.on taken. ~.                 ,-tSJ
        ~    ,
SRI Q f i e d by Notet that allowU day' ~31 IS        J~IA;:J          ~~
days <Eespeptivel'V after withdrawal of the control rod and increaslng power to above the LPSP, to perform the Surveillance. This acknowledges that the control rod must be first withdrawn and THERMAL POWER must be increased to above the LPSP before performance of the Surveillance, and therefore, the NoteJ avoid Potential conflicts with SR 3.0.3 and SR 3.0.4.                 ~1 SR    3.1.3.4 Verifying that the scram time for each control rod to 90%
insertion is ~ 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and (continued)
Dresden 2 and 3                      B 3.1.3-8                            Revision 0

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
LaSalle County Station, Units 1 and 2 Technical Specification Bases Pages B 3.1.3-4 B 3.1.3-8

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS         A.l. A.2. A.3. and A.4         (continued) control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if: a) the stuck control rod occupies a location adjacent to two "slow" control rods, b) the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pair of "slow" control rods elsewhere in the core adjacent to one another. The description of "slow" control rods is provided in LCO 3.1.4, "Control Rod Scram Times."
In addition, the associated control rod drive must be disarmed within 2 hours. The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable amount of time to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CROM or reactor internals. The control rod isolation method should also ensure cooling water to the CRO is maintained.
Monitoring of the insertion capability for each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.                     ~
SR 3. 1 3. 2 and' SR 3. 1. 3 . 3 per for rrfI)e rio di c t est s 0 f t he~
contro ro insertion capability of withdrawn control rods.
Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The allowed Completion Time provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.
(continued)
LaSalle 1 and 2                    B 3.1.3-4                                       Revision  a

Control Rod OPERABILITY B 3.1.3 BASES 3.1.3.   ~
SURVEILLANCE REQUIREMENTS (continued)   Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
The control rod may then be returned to its Original~~
po sit ion. This ens ures the con t r 01. rod i s no t stu c ~ and is*;-f IS:
free to insert on a scram signal. ~ c e a e .
not required when THERMAL POWER is less than or equal to the de&sect;)
actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of the analyzed rod position se uence (LCO 3.1.6) and the RWM (LCO 3.3.2.1).
The 7 day Frequ ncy of SR               1S ase on opera ng experience rel ted to the change in CRD performanc and the ease of perfor ing notch testin for fully withdra n control rods. Partia ly withdrawn con rol rods are teste at a 31 day Freque cy, based on the potential power re uction required to llow the control rod movement, and onsidering the lar e t   ting sam le of S 3.1.3.2. Furthe more, the
      ~'1H&#xa3;      31 day Frequency takes into account operat1ng experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.~                       ~
        ~~sRie  ))..        . ~              A modified by Note! that all ow 7 day* an 31 days es'pe ively, after withdrawal of the control rod and increasing power to above the LPSP, to perform the Surveillance. This acknowledges that the control rod must be first withdrawn and THERMAL POWER must be increased to above the LPSP before performance of the Surveillance, and therefore, the Notel avoi~-:otential conflicts with SR 3.0.3 and SR 3.0.4.             ~
(continued)
LaSalle 1 and 2                    B 3.1.3-8                               Revision 0

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
Oyster Creek Generating Station Technical Specification Bases Page 4.2-4

MONTHLY The we 1 control rod exercise test serves as a periodic check against deterioration of the control rod s stem. Experience with this control rod s ste ~indicated that we 1 tests are adequate, and that rods which move by drive pressure wi scram when required as the pressure applied is much higher. The requirement to exercise the control rods within 24 hours of a condition with two or more control rods which are ANt> AtJAL'IStS               valved out of service or one fully or partially withdrawn control rod PPlFdttM EY') ':::'Ot'L        wh i ch can not be moved provi des assurance of the re1i abi 1i ty of the AfJ ,rJDV$,{L:y-U.HOf          remai ni ng ~ontro 1 rods.
Be...uf<.. INI T/~TIV~
_            ppac\k-()   Pump operabil ity, boron concentration~ solution temperature and volume THA \      U;,A'fj;I .   'Iff   of standby 1iqu id control system 4 are checked on a frequency GY NfL()            11ft        consistent with instrumentation checks described in.Specification 4.1.
Experien~e wIth similar systems has indicated that the test frequencies are adequate. The only practical time to functionally test the liquid control system is duriJtg a refueling outage. The functional test includes the firing of explosive charges to open the shear plug valves and the pumping of demineral ized water into the reactor to assure operabi 1i ty of the system downstream of the pumps. The test also includes recirculation of liquid control solution to and from the' solution tanks.
Pump operability is demonstrated on a more frequent basis. This test consists of recirculation of demineralized water to a test tank. A continuify check of the firing circuit on the shear plug valves is provided by pilot lights in the control room.                                     Tank level and temperature alarms are provided to alert the operator to off-normal conditions.
Figure 3.2.1 was revised to reflect the minimum and maximum weight percent of sodium pentaborate solution, and the minimum atom percent of B-I0 to meet 10 CFR 50.62(c)(4). Since the weight percent of sodium pentaborate can change with water makeup or water evaporation, frequ~nt surveillances are perfotmed on the solution concentration, volume and temperature. The sodium pentaborate is enriched with 8-10 at the chemical vendor's facility to meet the minimum atom percent.
Preshipment samples of batches are analyzed for B-I0 enrichment and veri fi ed by an independent 1aboratory pri or to shi pment to Oyster Creek. Since the 8-10 enrichment will not change while in storage or in the SlCS tank, the surveillance for 8-10 enrichment is performed on a 24 month interval. An additional requirement has been added to evaluate the solution's capability to meet the original design shutdown criteria whenever the Boron-l0 enrichment requirement is not met.
                              , The functional test and other surveillance on components, along with the monitoring instrumentation, gives a high reliability for standby liquid control system operability.
References -
(1) FDSAR, Volume II, Figure 111-5-11 (2) FDSAR, Volume I, Section VI-3 (3) FDSAR, Volume I, Section 111-5 and Volume II, Appendix B (4) FDSAR, Volume I Section VI-4                                                                        _
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                                                                      )
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OYSTER CREEK                ____~              ~                      -      en men     0.: 75, 124, 159, 172, 178

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
Peach Bottom Atomic Power Station, Unit 2 Technical Specification Bases Pages B 3.1-16 B 3.1-19 B 3.1-20

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS      A.I, A.2, A.3, and A.4 (continued) location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another.
The description of "slow" control rods is provided in LCO 3.1,4, "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours.
The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CROM. The control rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD.
Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER reate than the low power setpoint (LPSP) of the RWM. ~$
R 3.1 3.2 and R 3.1.3.3 perform(periodic tests of the ~
contro r~ insertion capability of withdrawn control rods.
Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The allowed Completion Time of 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the LPSP of the RWM provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.
To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SOM is also required within 72 hours. Should a DBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SOM demonstration may not be valid. The SOH must therefore be evaluated (by measurement or analysis) with the stuck control rod at its (continued)
PBAPS UNIT 2                   B 3.1-16                     Revision No. 2

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS      ~      (continued) inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.
SURVEILLANCE  SR  3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room.
              ~ 3.A.3.2/a~SR         3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
The control rod may then be returned to its original position. This ensures the control rod is not stuck and" JHIS free to insert on a scram signal. Th s Surveillance ~
not required when THERMAL POWER is ess than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the analyzed rod position sequence CLCO 3.1.6) and the RWM CLCO 3.3.2.1).
The 7 day Freq ency a               is based on oper ting experience rel ted to the chan es in CRD performa ce and the ease of perfo ming notch test"ng for fully withdr wn control rods. Parti ly withdrawn c trol rods are test d at a 31 day Frequ ncy, based on t e potential power eduction required to allow the contr 1 rod movement and onsidering the large t sting sample of SR 3.1.3.2. Furth rmore, the
       ~~H   31 day Frequency takes int accoun operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a (continued)
PBAPS UNIT 2                     B 3.1-19                    Revision No. 63

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE (jR 73. 1.3/2 a~SR      3.1.3.3   (continued)
REQUIREMENTS determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.
For example, the unavailability of the Reactor Manual Control System does not affect th PERABILITY of the control rods, provided R 3 1.3 an SR 3.1.3.3~acurrent in accordance with SR 3.0*.
S SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is ~ 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or trans ient{ thereby comp 1et i ng its shutdown funct ion.
This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation," and the functional testing of SOV vent and drain valves in LCD 3.1.8, *Scram Discharge Volume (SOV)
Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
SR 3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CROM and will perform its intended function when necessary. The Surveillance requires verifying a control rod does not go to the withdrawn overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRO can reach the overtravel position.
The verification is reqUired to be performed any time a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or eRO System that could affect coupling (CRO changeout and blade replacement or complete cell disassembly, i.e., guide tube removal). This includes control rods inserted one notch and then retur.ned (continued)
PBAPS UNIT 2                      B 3.1-20                        Revision No. 0

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
Peach Bottom Atomic Power Station, Unit 3 Technical Specification Bases Pages B-3.1-16 B 3.1-19 B 3.1-20

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS       A.l, A.2, A.3, and A.4 (continued) location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another.
The description of "slow" control rods is provided in lCO 3.1.4, "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours.
The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CROM. The control rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRO.
Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpo i nt (LPSPl of the RWM. ,/""?)
             ~: 3/l.3.2/arrd)SR 3.1.3.3 performfperiodic tests of the ~
ntrot rod insertion capability of withdrawn control rods.
Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual lPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (lCO .3.1.6) and the RWM (lCO 3.3.2.1). The allowed Completion Time of 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the LPSP of the RWM provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.
To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SOM is also required within 72 hours. Should a OBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SOM demonstration may not be valid. The SOM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its (continued)
PBAPS UNIT 3                    B 3.1-16                      Revision No.2

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS          ~      (continued) inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.
SURVEI LLA NCE   SR   3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the~trol r~oom.....                       ~
                ~                                 .~ 3 .. 1.. 3 .. ~ DElt:,i=l)
                      'J/      /' .~
                @ 1'1.3.~aqE:JSR         3.1.3.3 Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
The control rod may then be returned to its original position. This ensures the control rod is not stuck an is H\S free to insert on a scram signal. h se Surveillance ~
not required when THERMAL POWER is less than or equal tCllthe ~
actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the analyzed rod osition se en e (LCO 3.1.6) and the RWM (LCO 3.3.2.1).
The 7 day Freq ncy of SR 3.1.3 2 is based on o~ rating experience rel ted to the chan s in CRD perfor ance and the ease of perfor ing notch testi g for fully wit drawn control rods. Partia ly withdrawn co rol rods are t ted at a 31 day Frequ cy, based on th potential powe reduction required to llow the control rod movement a ~ considering th           testing sam le of R 3.1.3.2. Fur hermore, the E[0 THE ~31 day Frequency takes into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a (continued)
PBAPS UNIT 3                         B 3.1-19                      Revision No. 64

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE ~.1.3;t2 ~SR 3.1.3.3 (continued)
REQUIREMENTS determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.
For example, the unavailability of the Reactor Manual Control System does not fect the OPERABILITY of the control rods, provided,SR 3 1.3.2 and SR 3.1.3.3~. e curre.nt in accordance with SR 3.0.2.
fS  I SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is ~ 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.
This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS)
Instrumentation,* and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV)
Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
SR 3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The Surveillance requires verifying a control rod does not go to the withdrawn overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position.
The verification is required to be *performed any time a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling (CRD changeout and blade replacement or complete cell disassembly, i.e., guide tube removal). This includes control rods inserted one notch and then returned (continued)
PBAPS UNIT 3                    B 3.1-20                      Revision No. 0

TECHNICAL SPECIFICATION BASES PAGES (Mark-ups)
Quad Cities Nuclear Power Station, Units 1 and 2 Technical Specification Bases Pages B 3.1.3-4 B 3.1.3-8

Control Rod OPERABILITY B 3.1.3 BASES ACTIONS           A.l. A.2. A.3. and A.4   (continued)
RWM is bypassed to ensure compliance with the CRDA analysis.
With one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The stuck control rod separation criteria are not met if: a) the stuck control rod occupies a location adjacent to two "slow" control rods, b) the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pai r of "slow" control rods el sewhere in the core adjacent to one another. The description of "slow" control rods is provided in LCO 3.1.4 "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours. The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CRDM or reactor internals. The control rod isolation method should also ensure cooling water to the CRD is maintained.
Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.
(SR3.1A.2 aYjSDSR 3.1.3.3 perforrrfperiodic tests of the @
control rod insertion capability of withdrawn control rods.
Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and (continued)
Quad Cities 1 and 2                B 3.1.3-4                         Revision 0

Control Rod OPERABILITY
_-----                    B 3.1. 3 BASES c-------tc:&sect;!'                     I. :5. d.. D eu:l~
SURVEILLANCE                            3.1.3.3 REQUIREMENTS (continued)    Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
The control rod may then be returned to its original position. This ensures the control rod is n t stuck and is ~HI&#xa3; free to insert on a scram signal.      h se Surveillance~~       l' not required when THERMAL POWER is less than or equal to the ~
actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the analyzed rod position sequence (LCO 3.1.6) and the RWM (LCO 3.3.2.1).
The 7 day Freq ency of SR .1.3. is based on operat"ng experience re ated to the change in CRD performanc and the ease of perfo ming notch testin for fully withdraw control rods. Parti lly withdrawn cont 01 rods are tested at a 31 day Freq ency, based on the potential power red ction required to allow the control od movement and co sidering the large esting sample of S 3.1.3.2. Further re, the day Frequency ta es into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.     .'              ~
        ~sRI a e modified b~1 that allow(! dab a@31 r                             I~
days respe tively after withdrawal of the control rod and increaslng power to above the LPSP, to perform the Surveillance. This acknowledges that the control rod must be first withdrawn and THERMAL POWER must be increased to above the LPSP before performance of the Surveillance, and therefore, the Note, avoi~. Protential conflicts with SR 3.0.3 and SR 3.0.4.             ~
SR   3.1.3.4 Verifying that the scram time for each control rod to 90%
insertion is ~ 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS)
(continued)
Quad Cities 1 and 2                 B 3.1.3-8                          Revision 0

ATTACHMENT 4 List of Regulatory Commitments Application for Technical Specification Change Regarding Revision of Control Rod Notch Surveillance Test Frequency, Clarification of SRM Insert Control Rod Action, and Clarification of a Frequency Example Using the The following table identifies those actions committed to by Exelon Generation Company, LLC (EGC) and AmerGen Energy Company, LLC (AmerGen) in this document. This commitment applies to Clinton Power Station, Unit 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2. Any other statements in the submittal are provided for information purposes and are not considered to be regulatory commitments.
COMMITMENT TYPE COMMITTED COMMITMENT                                    ONE-TIME ACTION    PROGRAMMATIC DATE (Yes/No)           (Yes/No)
EGC and AmerGen will establish the Technical Specifications Bases for TS Bases 3.1.3 and 3.3.1.2 consistent with those shown in TSTF-475, Revision 1, Implement with IIControl Rod Notch Testing                                    Yes                 No amendment Frequency and SRM Insert Control Rod Action. II [Note:
Oyster Creek Nuclear Generating Station TS Bases differ from the STS]
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