ML20113A677
| ML20113A677 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 12/31/1995 |
| From: | Mitchell T PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9606250145 | |
| Download: ML20113A677 (21) | |
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. f* Thomss N.Mitch1H Vre President ;
Peach Bottom Atomic Power Station
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PECO NUCLEAR ;=;g,;-ne"v l A Unit of PECO Energy P D,*"is^ N 7 l Fax 717 456 4243 June 20,1996 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555
SUBJECT:
Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3 Annual 10 CFR 50.59 Report For The Period January 01,1995 through December 31,1995 Gentleman:
Enclosed is the 1995 Annual 10 CFR 50.59 Report as required by 10 CFR 50.59 (b). If you have any questions or require additional information, please contact us. ;
1 Sincerely, '
Thomas N. Mitchell Vice President Peac m Atomic Power Station TM/ /GAJ 4 Attachment 4 cc: B. Gorman, Public Service Electric & Gas R. R. Janati, Commonwealth of Pennsylvania l T. T. Martin, US NRC, Administrator, Region 1
- R. l. McLean, State of Maryland W. L Schmkit, US NRC, Senior Resident inspector A. F. Kirby lli, DelMarVa Power H. C. Schwemm, VP - Atlantic Electric 240119 ,
CCN 96-14053 l
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R PDR a
4 Docket Nos.50-277 50-278 1995 PEACH BOTTOM ATOMIC POWER STATION ANNUAL 10 CFR 50.59 REPORT This report is issued pursuant to reporting requirements for Peach Bottom Atomic Power Station Units 2 and 3 (Faculty License Numbers DPR-44 and DPR-56 respectively). This report addresses tests and changes to the facDity and procedures as they are described in the Peach Bottom Final Safety Analysis Report. This report consists of those tests and changes that were implemented between January 1,1995 and December 31,1995. A safety evaluation for each item has concluded that no unreviewed safety questions, as defined in 10 CFR 50.59 (a) (2), were involved.
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CCN 96-14053
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1 PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWER STATION
- UNIT 2 AND 3 l i DOCKET NOS. 50-277 AND 50-278 '
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ANNUAL 10 CFR 50.59 REPORT 1 l
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j JANUARY 1,1995 THROUGH DECEMBER 31,1995 '
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l CCN 96-14053
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l PECO ENERGY COMPANY PEACH BOTTOM ATOMIC POWEH STATION UNIT 2 AND 3 DOCKET NOS. 50-277 AND 50-278 ANNUAL 10 CFR 50.59 REPORT TABLE OF CONTENTS i Enaineerina Channa il'M th;. . . . ............................ Page ECR 94-1 1398 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1
ECR 95-02243 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l ECR 95-05129 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 ECR PB95-01614 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 EWR A0660658 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 EWR A0913559 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Miscellaneous Safety Evaluations CAD Valve Position Change . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Change Commitment on NUREG 737 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1
Core Design Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 Core Flow Less than Natural Circulation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Core Operating Limits Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Emergency Preparedness Action Levels Tech. Basis . . . . . . . . . . . . . . . . . . . . . . . 3 Emergency Load Centers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1
Feedwater Heater Removal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Fire Protection Program Assumption of LOOP with a Fire . . . . . . . . . . . . . . . . . . . . 4 Mn Steam Bypass Valves (2) 00S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Operations 12 Hour Shifts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 l Power R erat e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Reactor Cavity Shield Plugs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 Recirc Pump Flow Adjustments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4
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,i Modifications M O D 2254 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M O D 5183 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M O D 53 70 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4
4 M O D 53 75 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 i.
MOD 5398 (ECR 9410461) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 e M O D 5400 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
+
MOD P000061 ...................................................6 4
MO D P000068 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 a
a MOD P000105 . . . . . . . .. .......................................7 i
4 M OD P000190 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l
i MOD P000207 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 MOD P000231 ...................................................7 MOD P000236 (ECR 9410005 & 11628) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 MOD P000262 (5414) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 MOD P000287 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 M O D P000309 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 M OD P000335 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 MOD P000444 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 MOD P0004 79 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 MOD P000496 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 CCN 96-14053
Non-Cor40nnence Reoorts Non-Conformance Report 9540027 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Non-Conformance Report 95-001 1 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Non-Conformance Report 95-041 16 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Non-Conformance Report P000033 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Plant Enhancement Prooram PEP 10004676 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Procedures A-C-023.......................................................11 CH-139.......................................................11 EP-C-003 & EP-C-003-1 ........................................... 11 ER P- 101 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 E R P.101 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 Emergency Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 MAT-5414N, O, & J (P000262) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Offsite Dose Calculation Manuals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 RT-F-037-31 1 -2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 SO-05.2. B-2 & GP-05 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 S0-10.1.A-3 TC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 S P-2098 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 Technical Requirements Manual . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 CCN 96-14053
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PECO ENERGY COMPANY !
PEACH BOTTOM ATOMIC POWER STATION j UNIT 2 AND 3 1 DOCKET NOS. 50-277; 50-278 ANNUAL 10 CFR 50.59 REPORT JANUARY 1,1995 THROUGH DECEMBER 31,1995 SAFETY EVALUATION SUMMARIES CCN 96-14053
.. j PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT
. ECR 94-11398 Year implemented: j
! Unit 2 (1995) I 1
Unit 3 (1995) !
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l This review evaluat wi the impact of the change in the High Pressure Coolant injection (HPCI) and Reactor i Core Isolation Cooling (RCIC) analytical response times. This change affected the UFSAR which specifies j HPCI and RCIC response times. The system stabuity and govemor tuning were not changed by this l i evaluatior' and no adverse safety concems were created as a result of this activity. Based on the Safety !
j Evaluation, it was determined that tMs change did not constitute an Unreviewed Safety Question. i
! ECR 95-02243 Year implemented:
! Unit 2 (1995) i Unit 3 (1995)
( The evaluation reviewed changes to the UFSAR Section 7.1.7. This change deleted the requirement that the
! method of identifying Reactor Protection System and Engineered Safety Feature equipment be a " Yellow j Nameplate". This change does not affect system operations during normal and emergency operations. No l adverse safety concerns were created as a result of this clunge. Based on the Safety Evaluation, it was i determined that this change did not constitute an Unreviewed Safety Question.
l ECR 95-05129 Year implemented:
l Unit 2 (1395) j Unit 3 (1995) l This activity temporarHy changed the position of a Containment Attwspheric Dilution (CAD) valve from normally opened to closed. This was done to prevent the loss of CAD Tank inventory through a relief valve.
This change is being done untu the issue is permanently resolved. This change affected the valves position as described in the UFSAR. No new transients, events or safety concems were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
! ECR PB95-01614 Year implemented:
j Unit 2 (1995)
Unit 3 (1995) ;
i i This activity allowed a control rod with a faulty reed switch at position '48' ta be fuHy withdrawn with no rod l
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j drift. This was done to clear the rod drift alarm so that the rod could be fully withdrawn. This change affected l control rod descriptions as specified in the UFSAR. No new type accidents or adverse safety concems were
! created as a result of this change. Based on the Safety Evaluation, it was determined that this change did j not constitute an Unreviewed Safety Question.
l l EWR A0660658 Year implemented:
j Unit 2 (1995)
- Unit 3 (1995) i i This evaluation reviewed and justified the use of lead shielding to reduce radiological fields around the I Control Rod Drive Hydraulic Control Units while the unit was at power. This change did not adversely affect i plant safety or create any new accident initiators or equipment malfunctions. Based on the Safety Evaluation.
It was determined that this change did not constitute an Unreviewed Safety Question.
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PEACH BOTTOM ATOMIC POWER STATION l UNIT 2 & 3
- Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT EWR A0913559 Year Implemerned
Unit 2 (1995)
- Unit 3 (1995)
This evaluation reviewed maintaining a constant reactor steam dome pressure of 1005 psig by periodically increasing turbine pressure setpoint during cycle 10 power coast < lown. This change affected the generic coast <f own analysis referenced in the UFSAR. No new transients, accidents or adverse safety concerns were q created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
1 CAD Valve Position Chance Year Implemented:
Unit 2 (1995)
- Unit 3 (1995)
This evaluation changed the position of a Containment Atmospheric Dilution (CAD) system valve from normally closed to the open position. This was done to allow the CAD sysum to be used for routine makeup during normal plant operations. This change affected the valves pos! !ca and system use as indicated in the UFSAR. No new accident, transients, malfunctions or adverse safety concems were created as result of this 2 change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
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Chanae Commitment on NUREG 737 Year Implemented:
Unit 2 (1995)
Unit 3 (1995)
This evaluation justified that the performance of component verifications can be performed by trained and qualified individuals who are not part of operations. This affected a response to NUREG 737 which identified Licensed and Qualified Floor operators as individuals able to perform verif' cations. The use of trained and 1 qualified non<>perations personnel to perform these verifications does not adversely affect the safety of the plant. b ised on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
2 Core Desian Reoort Year Implemented:
Unit 2 (N/A)
Unit 3 (1995)
, This evaluation addressed the Unit 3 Core Design Report for cycle 11 operations. The core load was designed to be compatible with existing fuel in the reactor. There was no impact on safety or increased probability of a failure as a result of this activity. Based on the Safety Evaluation, it was determined that this
- change did not constitute an Unreviewed Safety Question.
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i PEACH BOTTOM ATOMIC POWER STATION l UNIT 2 & 3 )
d Docket No. 50-277 & 50-278 ;
i 199510 CFR 50.59 REPORT ;
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- Core Flow Less than Natural Circulation Year implemented
i Unk 2 (1995) l Unit 3 (1995)
{ This evaluation reviewed operations of the plant with core flows less than the Natural Circulation Une as j indicated on the power to flow map. Plant operations in this region are described in the UFSAR. The core
, thermal limits identified in the Core Operating Umit Reports are sufficient to maintain the margin of safety
- in the newly defined region of the power / flow map. Based on the Safety Evaluation, it was determined that
- this change did not constitute an Unreviewed Safety Question.
j Core Operating Umits Report Year implemented:
j Unit 2 (1995) i Unit 3 (1995)
! This evaluation was performed to address changes to the Units 2 & 3 Core Operating Umit Reports. The 4
revision was made to allow for less restrictiv9 thermal limits (ARTS multipliers). The change will allow for
! additional operating thermal margins. This aciMiy affected documentation addressed in the SAR. In l addition, the change did not adversely affect any se.fety limits or operating modes. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
- Core Operatino Limits Report Year implemented
! Unit 2 (1995)
UnN 3 (1995)
! This evaluation was performed to address changes to the Units 2 & 3 Core Operating Umit Reports. The
! revision was made to incorpw.ie changes to the ARTS-based power dependent MCPR limits to correct an i error which introduced non-conservatism in the original cycle specific limits. This change aflected the COLRs 4
which are part of the SAR. No safety concems, accident initiator or mitigators were created. Based on the i Safety Evaluation, k was determined that this change did not constitute an Unreviewed Safety Question.
j Emeroencv Plan Action Levels Tech. Basis Year implemented:
! Unit 2 (1995) l Unit 3 (1995) i
! This evaluation reviewed the Technical Basis Manual which is used to develop emergency action levels
, different from those described in the Nuclear Emergency Plan and ERP 101. A revision tc, the plan and i ERP-101 is required to implement this change. The Emergency Preparedness Plan is considared part of the SAR. No adverse safety concems were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
l Emeroency Load Centers Year implemented:
{ Unit 2 (N/A)
{ Unit 3 (1995) 1 This evaluation justified the removal of control power fuses associated with the E-313 and E-333 degraded i voltage protective circuits and was done to support implementation of modification 5414 (P00262). This i temporary activity modified the 4 Kv emergency switchgear transfer logic as desc6 bed in the UFSAR. No j adverse safety concems were created as a result c4 this change. Based on the Safety Evaluation, it was l j determined that this change did not constitute an Unreviewed Safety Question. ;
Page 3 of 14 CCN 95-14053
. PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT Feed Water Heater Removal Year implemented:
Unit 2 (1995)
Unit 3 (1995) I This evaluation justified the operation of PBAPS Unit 3 during cycle 10 with a final feedwater temperature reduction of up to 55 degrees F during coast-down operations with the 5A, SB, SC and 4B Feedwater heaters removed from service. This affected plant and system operations as described in the UFSAR. No adverse temperature issues or transients were created as a result of this change. Based on the Safoty Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Fire Protection Proaram Assumotion of LOOP with Fire Year implemented:
Unit 2 (1995)
Unit 3 (1995)
This evaluation reviewed the safe shutdown analysis which assumes that Off-Site Power is lost. This change allowed taking credit for Off-Site Power when its loss is not fire induced and when altemative shutdown is not used. This change affected documentation specified within the Fire Protection Program. No adverse safety consequences were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Mn Steam Bvoass Valves (2) OOS Year implemented:
Unit 2 (1995)
Unit 3 (N/A)
This evaluation supports the continued operation of Unit 2 during cycle 11 with up to two Main Turbine Bypass Valves out of service. The UFSAR states that nine valves are available to support bypassing steam. '
This scenario is bounded by the generator load reject without bypass event. No adverse safety concems, 1
reldents or transients were created as a result of this change. Based on the Safety Evaluation, it was detenained that this change did not constitute an Unreviewed Safety Question.
, Ooerations 12 Hour Shifts Year Implemented:
Unit 2 (1995)
Unit 3 (1995) i i
This evaluation reviewed and approved the transition from 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shifts to 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> v ifts for Operations personnel. This change affected a NRC Safety Evaluation Report associated with the Ptioch Bottom Restart Plan. No adverse safety concerns were created as a result of this change. Based on the Safety Evaluation, R was determined that this change did not constitute an Unreviewed Safety Question.
Power Rerate Year Implemented:
Unit 2 (1994)
- Unit 3 (1995)
This evaluation justified the revision of a GE Nuclear Energy document Power Rerate Licensing Report with i
respect to ARTS. The change affected documentation specified in the SAR. The activity did not adversely affect plant safety or any opecting modes or accidents. Based on the Safety Evaluation, it was de. ermined that this change did not constitute an Unreviewed Safety Question.
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. PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT Reactor Cavity Shleid Pluas Year implemented:
Unit 2 ( )
Unit 3 (1995)
This evaluation justified the installation of the Reactor Cavity Shield plugs during reactor startup ep to a specified power level. The safety evaluation concluded that the radiological effects and consequences of an ,
accident have been analyzed without any adverse results. No new accident, transient or operating modes l are created as a result of this change. Based on the Safety Evaluation, it was oeim.!ned that this change did not constitute an Unreviewed Safety Question.
Recire Pumo Flow Adiustments Year Implemented: l Unit 2 (N/A) l l
Unit 3 (1995)
This evaluation provided justification to support continued operation with a Recirculating (Recirc) Pump Motor / Generator (M/G) set locked up during a plant transient. This activity affected how the Recirc M/G set system is described in the UFSAR. This change did not create any adverse safety concems by not tripping the Recire M/G set. Based on the Safety Evaluation, it was determined that this change did not i constitute an Unreviewed Safety Question. j MOD 2254 Year implemented:
Unit 2 (1995)
Unit 3 (1995) ;
J This modification installed a third Off-Site electrical power source and associated switchgear components. l This activity acts as a connectable spare from the other substation, replaces either of the existing preferred sources if unavailable and allows room for future load growths. This change affected documentation addressed in the UFSAR and did not adversely affect plant safety. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
i MOD 5183 Year implemented:
Unit 2 (1995)
Unit 3 (1995)
This modification addressed the refurbishment of the Unit 2 and 3 Reactor Building Cranes. The upgrade affected UFSAR section 10.4 descriptions. No new accidents, operating modes or off-normal conditions was created as a result of this activity. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD 5370 Year Implemented:
Unit 2 (1994)
Unl* O (1995)
This modification replaced the three existing 50 percent capachy pumps with two 100 percent capacity pumps and upgraded the existing filter demineralizer precoat cycle control circuitry and manual isolation valves. Tr:Is modification improves the Reactor Water Cleanup System by providing better overall system performance. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Page 5 of 14 CCN 96-14053
i PEACH BOTTOM ATOMIC POWER STATION i UNIT 2 & 3 J Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT l
1 MOD 5375 Year implemented: l Unit 2 (1994) ;
Unit 3 (1995)
This modification enhanced several alarm annunciator logic circuits to allow no alarms to be up during normal operations. Five alarm annunciators were normally up during power operations. This change allowed i the alarms to be cleared during power operations. This was done per INPO Good Practice. This affected l
a logic figure in the UFSAR. This change does not create any adverse operation conditions or adverse safety I concems. Based on the Safety Evaluation, k was determined that this change did not constkute an i Unreviewed Safety Question.
l MOD 5398 (ECR 9410461) Year implemented:
l Unit 2 (1995) '
Unk 3 (1995)
This modification replaced the existing plant process computer software (P1) with an enhanced type. This change affected the system descriptior; as specified in the UFSAR. No new accident, operating modes or transients have been created as a resJit of this actMty and k wNi not change any consequence previously considered. Based on the Safety Evaluation, it was determined that this change did not constkute an Unreviewed Safety Question.
MOD 5400 Year implemented Unit 2 (1995)
Unit 3 (1995)
This modification established a program for adding vibration sensing probes to rotating machinery and installing signal conditioner modules with protective enclosures, test station boxes and interconnecting cables. This actMty affected documentation addressed in the UFSAR for the associated systems. No adverse safety concerns were created as a result of the modification. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD P000061 Year Implemented:
Unit 2 ( )
Unit 3 (1995)
This modification replaced the existing Off-Gas system hydrogen analyzers and moisture monitor systems with upgraded units. This change affected the system as described in the UFSAR. This change did not adversely affect plant safety, any operating modes, or the station's abilty to monitor. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD PMMiB Year Implemented:
Unit 2 ( )
Unit 3 (1995)
This modification replaced the existing five Traverse Incore Probe (TIP) units with three new TIP units to increase system reliablity. This change affected the system description as specified in the SAR. No adverse safety concems were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Page 6 of 14 CCN 96-14053
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DEAC*i BOTTOM ATOMIC POWER STATION !
UNIT 2 & 3 l D%ket No. 50-277 & 50-278
- 199510 CFR 50.59 REPORT MOD P000105 Year Implemented
Unt 2 ( )
l Unit 3 (1995) l !
i This modification replaced the Main Steam, High Pressure Coolant Injection and Reactor Core isolation Cooling systems steam leak detection devices with new type digital units. This change affected the system's descriptions as specified in the UFSAR. The new type contrdiers have been evaluated to be equal or better
, than the previous units and no adverse safety concems were created by this modification, ,
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MOD P000190 Year implemented:
} Unit 2 (1994) i Unit 3 (1995) !
This modification replaced seven valve motor actuators with larger type units. This was done to meet
{ increased thrust demands requirements. The change affected documentation addressed in the UFSAR. No
- adverse safety concerns were created ouring this activity. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD P000207 Year Implemented:
l Urlt 2 (1994)
Unit 3 (1995)
- This modification changed the Feedwater / Recirculation Flow Contrd system instrumentation, alarms and runback logics. The chege affected documentation specified in UFSAR section 7.9. The activity did not i adversely affect plant operations, ability to response to transients or accident analysis. Based on the Safety i Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question. !
- MOD P000231 Year Implemented:
i Unit 2 (1995)
Unit 3 (1995)
This modification provided several enhancements for the Emergency Diesel Generators and support systems. The enhancements included relay upgrades, installation of a Data Acquisition System and associated computer connections, control logic changes to support testing and several support system improvements. These changes affect the diesels and support systems as described within the UFSAR. No adverse safety consequences or new transient, accidents or malfunction of equipment important to safety vare created as a result of this modification. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD P000236 (ECR 9410005 & 11628) Year implemented:
Unit 2 (1995)
Unit 3 (1995)
This modification replaced and relocated the High Pressure Coolant injection system Govemor Control box.
In addition, it also relocated the starters associated with the Auxiliary OH Pump, Condensate Pump and Vacuum Pump. These change impact component locations as specified in the UFSAR. However, these controls and relays have been relocated to a better environment. System operatiom vere maintained the same and no adverse safety concems were created as a result of this charge. Based on the Safety Evaluation, it was determined that this change did not constitute an Unrekt ad Safety Question.
Page 7 of 14 CCN 96-14053 F
PEACH BOTTOM ATOMIC POWER STATION
<- UNIT 2 & 3
) Docket No. 50-277 & 50-278-1 190E 10 CFR 50.59 REPORT MOD P000262 (5414) Year implemented:
i Unit 2 (1995) j Unit 3 (1995) i
) This modification changed the 4 Kv source breaker and auxNiary control logic to prevent repeated cycling i d the source breakers due to a combination of marginal grid voltage and a Loss of Coolant Accident signal
- i. In addition, the logic and the Core Spray logics were changed to prevent spurious tripping Furthermore, 3 the Emergency Diesel Generator Control circuits were changed to provide redundant start signals. These changes affected the systems as described in the UFSAR. No new accident modes or new type falures were
?
created as a result of this change. Based on the Safety Evaluation, it was determined that this change did
- not constitute an Unreviewed Saf ety Question.
i MOD P000287 Year implemented:
Unit 2 (1994)
Unit 3 (1995)
! This modification revised the reactor water level measurement pressure compensation to account for
- increased operating pressures and increased ambient temperatures in the drywell and reactor building. This i change affected documentation specified within the SAR. No adverse system operations or new concems i were created as a result of this activity. Based on the Safety Evaluation, k was determined that this change j did not constitute an Unreviewed Safety Question.
E l MOD P000309 Year implemented:
- Unit 2 (1995)
}
Unit 3 (1995)
! This modification replaced the old Diesel Driven Fire Pump Fuel Storage Tank (under ground) with an above
- ground tank. This was done to support more restrictive Ultrasonic Testing regulations. This change affected
)- tank location as described in the UFSAR. The diesel functions per its original design. Based on the Safety j Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
, MOD P000335 Year Implemented:
i Unit 2 ( ' )
} Unit 3 (1995) l Thls modification repaired weld cracks on the Unit 3 Core Spray line downcomers et the shroud penetration l location using clamping devices. The device ensures the structural integrity of the down comer is 2
maintained. This change, from a technical standpoint, did not involve a change to the Core Spray design basis as described in the SAR. However, for conservatism, a safety evaluation was performed since the Core Spray line is safety-related. No adverse safety concerns or new transients / accidents were created due to the implementation of the modification. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Page 8 of 14 CCN 96-14053 I
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l PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT MOD P000444 Year implemented:
Unit 2 ( )
Unit 3 (1995)
This modification installed bypass piping on several Residual Heat Removal, Core Spray, Reactor Core Isolation Cooling and High Pressure Coolant injection system valves to provide a continuous vent path from ;
the valve bonnet area to the downstream side of the valve. This vent path assures that the bonnet area de- '
pressurizes along with the associated piping. This change affected vah/e descriptions as specified in the UFSAR. Valve functions and isolation capability were not affected by this change and no adverse safety concerns were created as a result. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD P000479 Year Implemented:
Unit 2 ( )
Unit 3 (1995) ,
This modification replaced the Average Power Range Monitoring (APRM) flow blasing instrumentation with a Moore controller to enhance the reliability of the components. This change affected the system descriptions as specified in the UFSAR. The change has been evaluated and does not create any adverse safety concerns or new type accidents. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MOD P000496 Year implemented:
Unit 2 (1995)
Unit 3 (1995)
This modification replaced the Circulating Water intake Sampling System with a more reliable and self-contained composite sampling system. This change affected documentation and drawings listed in the UFSAR. The new system is better than the original system and no adverse safety concems were created.
Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Non-Conformance Reoort 95-00027 Year Implementec:
Unit 2 (1995)
Unit 3 (1995)
This Non-Conformance Report identified an improper assumption with respect to the loads ine',uded on UFSAR and Calculation tables. Specifically, miscellaneous loads associated with SU-25 Breaker Auxiliary Equipment. The correct loads were evaluated and did not adverse safety concems or new e'.:okients or transients were created as a result of this activity. Based on the Safety Evaluation, it was deterniined that this change did not constitute an Unreviewed Safety Question.
Page 9 of 14 CCN 96-14053
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PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199610 CFR 50.59 REPORT Non-Conformance Reoort 95-00111 Year implemented:
Unit 2 (N/A)
Unit 3 (1996)
This Non-Conformance Report addressed a repair to the Unit 3 High Pressure Service Water pump bay inist valve motor-operator. It was identified that cracks were present on two of the four motor-operator footers which hold the operator to ks base plate. This evaluation also justified revising the improved Technical Specification Bases 3.7.3 to clarify Emergency Heat Sink operability wkh one of the two gravity feed lines available. This change did not adversely affect plant safety. Based on the Safety Evaluation, k was
, determined that this change did not constitute an Unreviewed Safety Question.
.NQ!1-Conformance Reoort 95-04116 Year Implemented:
Unit 2 (N/A)
Unit 3 (1995)
This Safety Evaluation evaluated the use-as-is disposition of the Non-Conformance Report to allow the temporary removal of primary containment access shielding from the Control Rod Drive, personnel access and equipment hatches. The shielding is being removed to support refueling outage activities. The shielding being installed is described in the UFSAR section 12.0. Removal of the shielding does not adversely affect the environmental qualifications of surrounding equipment and does not create any new accidents, malfunctions or transients not described in the SAR. In addition, no safety concems were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Non-Conformance Reoort P000033 Year implemented:
Unit 2 (1995)
Unit 3 (1996)
This evaluation reviewed a potential non-conformance associated with electrical separation requirements specified in the UFSAR. The cables of concem are associated wkh the Main Control Room VentRation radiation monitor preampilfiers. Portions of these cables are not enclosed in conduit and do not meet ,
specified separation distances. Disposition of this evaluation was to use-as-Ic. The condition was reviewed i and does not create any adverse safety concems or the possible loss of any systems. Based on the Safety l Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
PEP 10004676 Year implemented:
Unit 2 (N/A)
Unit 3 (1995)
This evaluation justified the use of manual operations vice automatic operations of the Drywell Equipment and Floor Drain Sump Pumps. This change affected how the system is operated as described in the UFSAR.
System purpose has been reviewed to ensure that menual operations will adequately detect any identified or unidentified containment leakage. Based on the Safety Evaluation, k was determined that this change did not constitute an Unreviewed Safety Question.
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Page 10 of 14 CCN 96-14053
PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT MA23 Year implemented:
Unit 2 (1995)
Unit 3 (1995)
This procedure, which is ===elatad with Plant Evaluation & Special Test (PEST) Program was revised to remove two NRC commitments. These commitments were associated with the Technical Specification LCO entry mechanism and non4icensed individuals involved in manipulation of reactivity. This is acceptable based on these requirements being contained within other program procedures. No new accident initiators or adverse safety concerns were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
.CH-139 Year implemented:
Unit 2 (1995)
Unk 3 (1995)
This new procedure sets up and operates a temporary ISCO sampler associated with the Circulating Water, Water Treatment and Sewage Treatment systems. This activity was not consistent with the system sampling description as specified in the UFSAR. This system does not interface with any safety-related equipment and does not create any safety concoms. Based on the Safety Evaluation, k was determined that this change did not constitute an Unreview.. J Safety Question.
EP C-003 & EP-C-003-1 Year implemented:
Unit 2 (1995)
Unit 3 (1995)
These procedures provide the guidance for ensuring that adequate faculties and equipment to support emergency responses are provided and maintained consistent with the Nuclear Emergency Plan. The Nuclear Emergency Plan was affected wkh the implementation of these procedures. No adverse Emergency Preparedness or safety concem were created as a result of these changes Based on the Safety Evaluation, k was determined that this change did not constitute an Unreviewed Safety Question.
ERP-101 Year implemented:
Unit 2 (1995)
Unit 3 (1995)
The emergency response procedure was revised to address changes to specific action level classification.
t These changes were made due to the Vent and Main Stack Upgrade modification, tomados indications and
, river flooding. This activity affected SAR documents. No adverse safety concems were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constkute an Unreviewed Safety Question.
ERP 101 Year implemented Unit 2 (1995)
Unit 3 (1995)
The emergency response procedure was revised to address replacement of the Main Stack radiation monitoring syi, tem. This change affected the Emergency Plan. No new accident or transient initiators were created by this change and did not adversely affect any accident mitigators. Based on the Safety Evaluation, k was determined that this change did not constitute an Unreviewed Safety Question.
Page 11 of 14 CCN 96-14053
_ - _ . _ . _ _ _ _ _ _ . _ . _ . _ . . _ . _ _ _ . . _ . _ . _ _ _ _ - - - _ _ _ _ _ _ _ _ . ~ _ . _ _ __
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PEACH BOTTOM ATOMIC POWER STATION l UNIT 2 & 3 Docket No. 50-277 & 50-278 j 199510 CFR 50.59 REPORT l Emeroency Plan Year Implemented:
Unit 2 (1995)
Unit 3 (1995) l The Nuclear Emergency Plan changes are administrative in nature and were part of the 1995 annual review.
l The changes affected the facIky as described in the SAR because this plan is part of the SAR by reference.
No physical changes are being done and this activity did not create any adverse safety concem. Based on
- the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
MAT-5414N. O. & J (P000262i Year Implemented:
Unit 2 (1995) 4 Unit 3 (1995) 1 j This modification acceptance test verified that modification 5414, which installed enhancements to the 4 Kv 2 Emergency Switchgear logic, adequately testing the new design. This activity was considered a test not
- described in the SAR. No adverse conditions were created during the performance of this test. Based on the Safety Evaluation, k was determined that this change did not constitute an Unreviewed Safety Question.
, Off-Site Dose Calculation Manual Year implemented:
i Unit 2 (1995) 1 Unit 3 (1995)
- l j This evaluation reviewed and justified changing the PBAPS ODCM to reflect the current Radiological i
~
Environmental Monitoring Program (REMP). The changes involved discontinuing seven TLD stations to reduce the total number of sampling locations from 47 to 40 and replacing air particulate and air iodine sampling station 12 D with station 22G1 from LGS REMP. This change did not adversely affect any plant j safety or validity of the sampling systems. Based on tha Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
j Off-Site Dose Calculation Manual Year implemented:
- . Unit 2 (1995)
Unit 3 (1995)
- This change to the Off-Site Dose Calculation Manual addresses the implementation of the Main Stack and Vent Stack radiation monitor upgrade modification. The old flow paths were addressed in the ODCM and a needed to be revised to reflect the modification. This affected ODCM sections which are part of the SAR.
j No new accidents or mitigators were created as a result of this change. Based on the Safety Evaluation, it j was determined that this change did not constitute an Unreviewed Safety Question.
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Page 12 of 14 CCN 96-14053 4 - . - . , - . - .-_ -. - - .. , ,
. PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT
.RTf-037-311-2 Year implemented:
Unit 2 (1995)
Unit 3 (N/A)
This procedure performs a visual inspection of fire penetration seal's. The method of selecting the sample population of which penetrations wRI be inspected wil be changed. The activity allows the inspection of penetration seals without using a computer program to select those seals to be inspected as committed to the NRC. The Mainframe computer used to select the penetrations is no longer functioning so a new computer program will be used. This change wiu not adversely affect the reliabuity of the random sample but wRl change the tools used to identify the sample. Based on the Safety Evaluation, k was determined that this change did not constitute an Unreviewed Safety Question.
SO-05.2.B-2 & GP-05 Year Implemented:
Unit 2 (1995)
Unit 3 (N/A)
This evaluation revised GP-05 and S0-05.2.B-2 to allow power recovery to 100 percent power with a .'
Feedwater string out of service. This affected Feedwater string operation as described in the UFSAR.
Removal of one string does not create any adverse safety concems or any new type accidente or transients.
Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
S0-10.1.A-3 Year implemented:
Unit 2 (N/A)
Unit 3 (1995)
This procedure change allowed the operation of up to all three Fuel Pool FRter Domins on one unit to support cleanup of the cavity during Unit 3 refueling activities. This UFSAR indicates that one domin is line up to Unit 2, one is lined up to Unit 3 and the other is used when either of the other two demins are removed from service to support precoating. No new accidents, transients or adverse safety concems were created on this unit or the other unit. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
SP-2098 Year implemented:
Unit 2 (1995)
Unit 3 (1995)
This procedure manually valved out the APRM Flow Blas Transmitters during end-of-cycle coast-down. This affected the system description as specified in the UFSAR. Having the APRM flow bias signal valved out maintains the Rod Block and Scram setpoints stRI more conservative than those required by the Technical Specifications during end-of-cycle conditions. No adverse operating corw11tions or new accidents were created as a result of this change. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
Page 13 of 14 CCN 96-14053 i
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. PEACH BOTTOM ATOMIC POWER STATION UNIT 2 & 3 Docket No. 50-277 & 50-278 199510 CFR 50.59 REPORT Technical Reautrements Manual Year Implemented:
3 Unit 2 (1995)
- Unit 3 (1995)
This procedure change altered fire protection system requirements specified in the Technical Requirements Manual. This activity made several changes associated with the fire protection system which is required by a condition specified in the Facky Operating License. These changes did not adversely affect plant safety or create any new type malfunctions or accidents. Based on the Safety Evaluation, it was determined that this change did not constitute an Unreviewed Safety Question.
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i Page 14 of 14 CCN 96-14053