ML20012C651
| ML20012C651 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/31/1989 |
| From: | CLEVELAND ELECTRIC ILLUMINATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PY-CEI-NRR-1141, NUDOCS 9003230002 | |
| Download: ML20012C651 (288) | |
Text
e THE CLEVELAND ELECTRIC ILLUMINATING COMPANY P.O. SOX 07 5 PERRY. OHIO 44001 g TELEPHONE (216) 2643737 A ADDRESS 10 CENTER ROAD FROM CLEVELAND. 479 1200 9 TELtX: 2416s0 ANSWERsACK CEl PRYO Al Kaplan Serving The Best location in the Nation PERRY NUCLEAR POWER PLANT March 13, 1990 PY-CE1/NRR-1141 L U. S. Nucicar. Regulatory Commission Document Control Desk Washington, D.C. 20555 Perry Nuclear Power Plant Docket No. 50-440 Annual Report of 10 CFR 50.59 Safety Evaluations for 1989 Gentlement Attached is our summary report of 10 CFR 50.59 safety evaluations for the period of September 19, 1988 through September 18, 1989. An applicability check using the 10 CFR 50.59(a)(1) threshold critoria was performed on proposed changes to the design of the plant, to procedures / instructions, and to tests. All those
[ meeting the threshold criteria were further evaluated pursuant to the 10 CFR i 50.59(a)(2) criteria and are summarized herein.
This report summarizes a total of 306 safety evaluations, none of which resulted in the identification of an unreviewed safety question. Safety evaluations are numbered sequentially and those not included in this summary, have either been voided, withdrawn or are still under consideration though not approved at this time. Attachment 1 lists the safety evaluations divided into their major
' categories by type of item being evaluated. Attachment 2 defines the acronyms and format description. Attachment 3 provides the summaries of the safety evaluations described above.
Please feel free to call if you have any questions or comments.
Very truly y re, y- -
Al Kaplan Vice President Nuclear Group chments &
ce: T. Colburn [I P. Hiland 3 / f, g USNRC, Region III 9003230002 891231 g
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Summary of 1989 Perry Evaluations by Category '
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- The safety evaluations are divided into the major categories listed below.
Percentage Category Number of Total
- 1. Design Changes (except setpoint changes)' 30 10 s
2.' Drawing Changes 120 39
- 3. Setpoint~ Changes 7 2
- 4. USAR Changes 44 14
- 5. Procedure / Instruction Changes 43 14
-(revisions, temporary changes) ;
Lifted Lead & Jumper, Electrical Devices and
- 6. 17 6 l 1
Mechanical Foreign Item Changes i
- 7. Nonconformance Report' Evaluations- 22- 7
- 8. Special or Temporary Test Instruction Evaluations 12 4
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9.- Miscellaneous 11 4-Total 306 100 t
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Attech:3nt 2 PY-CEI/NRR-1141 L FORMAT DESCRIPTION Each 50.59 Safety Evaluation summary is presented in the following format SE No.: A sequentially assigned number from one (001) to end of h the period, preceded by the years e.g.86-025.
Source Document: There are several sources of evaluations which are abbreviated as shown.
DCN - Draving Change Notice DCP - Design Change Package EPI - Emergency Plan Instruction U FCR - Field Change Request FSAR CR - Final Snfety Analysis Report Change Request t
ISS - Installation Standard Specification--
LL&JED - Lifted Lead and Jumper and Electrical Device MFI - Mechanical Foreign Item NR-MM0X - Nonconformance Report; Maintenance & Modification Ouclity Section NR-NEDX'- Nonconformance Report; Nuclear Engineering Department NR-PPDX.- Nonconformance Report Perry Plant Department (s) where X - S or N., safety or nonsafety related PAP - Plant Administrative Procedure PEI - Plant Emergency Instruction PSP - Physical Security Plan SCR - Setpoint Change Request SOI - System Operating Instruction SSCR - Safe Shutdown Capability Report SVI - Surveillance Test Instruction ,
SXI - Special Test Instruction TAF - Technical Assignment File TXI - Temporary Test Instruction USAR CR - Updated Safety Analysis Report Change Request VMRN - Vendor Manual Revision Notice VO - Work Order Description of. Change:
A short narrative describing the location and type of plant change. For multiple i evaluations the discipline is identified in parenthesis for example, (Mechanical !
. Evaluation).
l Summary l
I. Response to 10 CFR 50.59(a)(2)(1) - is the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report increased?
II. Response to 10 CFR 50.59(a)(2)(ii) - is the possibility for an accident or l malfunction of a different type than any evaluated previously in the safety ,
analysis report created?
III. Response to 10 CFR 50.59(a)(2)(iii) - is the margin of safety as defined in the basis for any Technical Specification reduced?
.NJC/ CODED /3198
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-, ;l , . PY-CEI/NRR-1141 L P Page 1 of 285 u
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5 PERRY NUCLEAR POVER PLANT c,
SAFETY EVALUATION
SUMMARY
PURSUANT TO 10 CFR 50.59(b)(2) ,
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, . PY-CEI/NRR-1141 L Page 2 of 285-SE No.: 87-119, Rev. l~
Source Documents. LL&JED 1-87-097 Description of Change Lift all feedvater and condenser inputs to the reactor recirculation flov :
control valve circuitry so as to prevent flow control valve runbacks -
while' operating recirculation pumps at lov speed. ;
Summary
-I. No. 'The runback circuitry is not necessary during the operation of the recirculation pumps at lov speed and the lif ting of the subject i leads does not affect.any other plant circuitry. USAR Section 7.7.1.4.b.3 describes a runback as anticipatory.
l II. No.- Operation and control of the recirculation flow control valves will be maintained and runback is only required during high speed pump operation. No other circuits are affected.
III. No. Runback is only necessary for high speed operation of the recirculation pumps.
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, . PY-CEI/NRR-1141 L Page 3 of 285 E SE No.: 87-432 .
Source Document: SCR 1-87-1605 NR NEDS-2787 Description of Change This setpoint change revises the torque switch setpoint for valve 1B21-F0016 to be above the vendor's (Limitorque) recommended value.
Summary I.-No. Valve 1B21-F0016 is determined operable. Limitorque states that their operators can be set so that final thrust does not exceed 120%- -)
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of rated capacity, with some restrictions. The SMB-000 type operator used on this unit is rated for 8,000 lbs, thrust. Torque switch settings of 3.0 open and close vill deliver approximately l- 8,850 lbs of thr~ust. Limitorque states that if an operator is ,
overthrust to more than 110% but less than 120% of its published
- rating that the valve should be limited-to 100 cycles at which time an inspection of the components of the unit is performed to check for any cracking. If no cracking is found the operator can be returned to normal service at or belov vendor recommended settings.
By limiting operations to below 100 operations the probability of occurrence or the consequences of an accident or malfunction of 3 equipment previously evaluated in USAR is not increased. An NR has been generated to establish a PLC0 to ensure the valve is not
. operated in excess of 100 cycles without an inspection being conducted on the unit.
II. No. The slight overthrust condition for valve 1B21-F0016 vill not create the possibility of an accident'or malfunction different from any previously evaluated.
III. No. Operability of valve 1B21-F0016 as identified in Technical Specifications vill not be changed.
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AttS. chm 2nt 3 l PY-CEI/NRR-ll41 L Page 4 of 285-SE No.: 87-469 Source-Document: DCP 87-0450, Rev. O Description of Change '
This' design change installs a fuse / fuse holder and closes a circuit breaker to energize a nonsafety-related motor. space. heater for the Hydraulic Power Supply Unit (IF42-D002) to the Puel Transfer (F42) System preventing moisture from condensing on the motor internal parts when the pump is not running. Operating License Amendment 17 modified plant Technical Specifications prior to implementing this DCP.
Summary I. No. Closing of a circuit breaker and installation of a fuse to energize the existing nonsafety motor space heater preventing condensation on the motor internal parts does not change the function of the Hydraulic Power Supply Unit. Therefore, the probability of an accident or equipment malfunction previously evaluated is not increased.
II. No. Energization of the motor space heater (lF42-D002) vill not create y
the possibility of an accident or malfunction different from any previously evaluated.
l .III. No. Energization of the existing nonsafety motor space heater as designed does not impact the margin of safety, i
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. PY-CEI/NRR-1141 L Page 5 of 285
! .SE No.: .87-527; Rev. I source Document: DCP 87-0536, Rev. O Description of Change Revise the logic of the Reactor Core Isolation Cooling (RCIC) System
-turbine. trip throttle valve 1E51-F510 to reflect the revised opening cycle.. This' design change revires the valve to close and seat on torque, and revires limit switch LS-7 to provide proper indication.
Summary I. No. The RCIC turbine trip throttle valve 3E51-F510 acts as a quick closing, emergency trip valve to protect the turbine from damage upon receipt of a turbine trip signal.- The valve is used in the ll normal shutdown-of the system and has no active safety function for opening or closure using the motor operator. The change is required to protect the motor-operated valve from overtorquing. Therefore, the probability of an accident or equipment malfunction previously .i evaluated is not increased.
II. No.- The scope of the subject change is limited to the valve E51-F510 opening cycle logic modification described in Item I, above. As :
discussed previously, the design function of the RCIC trip / throttle valve has not been altered and required RCIC system performance .i requirements are maintained. In addition, no other systems or, components / equipment are affected by this change. A different type ,
of accident.or malfunction than previously evaluated in the USAR is '
thus not created.
III. No. Modification of the RCIC trip / throttle valve opening cycle logic has no impact on the design function or operability requirements of the RCIC system described in Technical Specification Sections 3/4.3.2, ;
3/4.3.5, 3/4.7.3, or associated bases. The margin of safety as .l defined for any Technical Specification basis is thus not reduced. j ll
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. PY-CEI/NRR-1141 L Page 6 of 285 SE No.- 88-401 '
Source Document PAP-0507, Rev. 7, TCN 3 '
Description of Change This safety evaluation evaluates changes made in PAP-0507 Preparation, Review and Approval of Instructions. This change updated the review and approval matrices:
o Review of Maintenance Section Instructions by maintenance unit supervisor o Preparation and approval responsibilities for Volume 10E-FTI's
" Refueling Instructions" vill be assumed by the Technical Section.
o Clarified the sections in the Nuclear Engineering Department responsible for Volume 4E-PSTG's " Perry Specific Technical Guides" l
o Corrected Training Manual titles and instruction numbers.
These instructions were formerly under one section and following.
reorganization these activities are now performed by several sections.
Summary I. No. The required instructions are being prepared by another qualified organization.
II. No. A change in responsibility for preparation of Fuel Technical Instructions does not create a nev' type of accident or malfunction.
III. No. The changes described do not affect the margin of safety defined in the Technical Specifications. '
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, , PY-CEI/NRR-1141 L Page 7 of 285-SE No.: 88-402 Source Document: DCP 87-0077, Rev. 0 Description of Change This design change adds a drain line from the High Pressure Core Spray (HPCS-E22B) Diesel Engine combustion air box drain connection to the floor drain, providing protection against excessive oi1~ build up in the air box without impacting diesel generator performance. A pressure indicator and isolation valves are being added to provide means of measuring air box pressure.
-Summary r
I. No. The probability of occurrence of a malfunction of equipment is reduced as the result of this design change. Opening the air box drain connection vill prevent the accumulation of oil in the HPCS diesel engine air box which would potentially occur during extended light load engine operation.
The ability to monitor air box pressure at full load, which is added by this change, enhances overall HPCS diesel engine reliability by providing additional engine performance information for the purpose of maintenance trending. The upstream isolation valve is provided .
for gage calibration purposes, such that the pressure indicator can be removed from service without affecting HPCS diesel engine operability.. The design codes and installation standards utilized to install this change are equivalent or identical to those used in the. original qualified design.
Since the overall reliability of the HPCS Diesel Generator.has been enhanced by this change, and the performance levels, design codes, l' and installation standards used in the original design have been l
maintained, the probability of occurrence of a malfunction of equipment important to safety evaluated previously is unchanged.
This design change creates no cross-ties between redundant components / systems. Therefore, in the event of an accident the HPCS Diesel Generator's response to mitigate its consequences remains unchanged, and the consequences of any accident evaluated previously remain unchanged.
Since the reliability, performance, and redundancy levels provided are equivalent to the original design, the response of the HPCS Diesel Generator to a malfunction of equipment important to safety is unchanged. Hence, the consequences of a malfunction of equipment are unchanged with respect to this change.
II. No. No new types of design or equipment are introduced by this design change. Since no new design type is introduced by this change, no l potential for an equipment malfunction of a different type than any I evaluated previously is introduced either.
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Attach::nt 3 PY-CEI/NRR-1141 L f Page 8 of 285 SE No.: 88-402 (Continued)
Summary III._No. The bases of Technical Specification Section 3/4-8 refer to the reliability of the onsite pover supply.- As demonstrated in Item I-above, the reliability of the HPCS Diesel Generator has been maintained with respect to this change, and the remaining divisional diesel generators are not affected-by this change. Since overall power supply reliability has been maintained, the margin of safety remains unchanged. No other Technical Specification is affected by thiu change.
'SE No.: 88-403 Source Document: DCP 88-0068, Rev. O Description of Change Install interlocks to automatically open the Emergency Service l Vater (ESV-P45) inlet and outlet valves (1P45-P014A,B and IP45-F068A,B) i to the Residual Heat Removal (RHR-E12) System Heat Exchangers when the associated Emergency Service Water Pump 1P45-C001A/B starts.
Summary I. No. This change adds circuits to the Safe Shutdown Capabilities Report (SSCR) which is part of the Approved Fire Protection Program j referenced in the USAR. However, this change vill not preclude safe. j shutdown during a fire nor does it alter the basis for any l Appendix R deviation request. Therefore, the probability of j occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. This change does not adversely affect the Fire Protection Program; I therefore, the possibility of an accident or-malfunction of a -i different type than evaluated in the USAR is not created.
- III. No. Only administrative aspects of the Fire Protection Program and Alternate Shutdown capability are defined in the Technical -;
Specifications. This change has no impact on the administrative !
aspects of the Fire Protection Program and does not adversely affect Alternate Shutdown capability. Therefore, the margins of safety as 1 defined in the bases for the portions of the Technical Specification applicable to fire protection are not affected.
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Page 9 of 285'
-SE No.: 88-404
' Source Document, DCP 88-0121A, Rev. O Description of Change r.
This change provides for the installation of a card reader door at door
[ location AX-403 to permit access to the new Health Physics Rear Access Building.
Summary I. No. The card reader door provides the same level of safety as the previously installed, non-card reader door.
II. No. Installation of this "A" label exterior door, provides the same level of safety as was previously provided by the non-card reader door, p III. No. Exterior doors are not specified in the Technical Specifications.
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, PY-CEI/NRR-1141 L Page 10 of 285 SE No.: 88-405 4
S6urce Document: DCP 88-0053, Rev. 0 '
Description'of Change i
The present configuration of the Hydrogen Analyzer (M51) System sampling lines preclude the Local Leak Rate Test (LLRT) on valves.1M51-F210A and B. This design change installs threaded capped tees in place of the existing elbows to serve as test connections for.LLRT testing.
Summary I. No. . This design change consists cf replacing an elbov vith a tee connection containing a 2-inch long nipple and threaded cap on each Hydrogen Analyzer containment sampling line. .The tee and threaded cap will~ serve as a test connection for LLRT purposes. The tees, nipples, and caps vill be procured and installed to the same l
' material and design specifications as the existing fittings, and ,
vill not decrease the structural performance of the line. In addition, the increase in line losses as a result of the tee is I negligible and vill not affect system flow rate. The tee vill be.
installed in such a way so as to prevent condensation build-up, thereby precluding the possibility of line blockage. Therefore, this design change vill not affect the ability of the Hydrogen Analyzers to perform their safety function of sampling and analyzing the Containment /Dryvell post accident atmospheres.
II.-No. See Item I above.
III. No. The addition of the test connections does not change the operation i of the M51 System or the plant as governed by the~ Technical Specifications. Thus the margin of safety as defined in the bases for the Technical Specifications has not been reduced.
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. . PY-CEI/NRR-1141 L s Page 11 of 285 SE No.: 88-406 Source Document: DCP 88-0302, Rev. 0 Description of Change This design change extended the Steam Jet Air Ejector (SJAE) loop seals by six feet and made several changes to the control and alarm functions-associated with the loop seals drain valves.
Summary I..No. The failure of the air ejector discharge lines to the Of fGas System were considered in the initial design of the plant. The failure is analyzed'in Section 15.7.1.3 of the.USAR. By lowering the lov level alarm level switch on the intercondenser loop seal this modification still meets the design specification for original installation without any additional equipment. Thus.the probability of occurrence or the consequences of an accident / malfunction of equipment'important to safety previously evaluated is not increased.
II. No. The piping modification vill not' change the original function of the Condenser Air Removal System. Therefore, the probability of an accident'or malfunction of a different type than any evaluated previously-in the USAR does not exist.
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- III..No. The Condenser Air Removal System still retains its original function with the modification of. the SJAE intercondenser loop seals. The piping modification vill not affect the operability or availability of the OffGas System (Base 3/4.11.2.4) or the Ventilation Exhaust- '
Treatment System (Base 3/4.11.2.5). The concentration of hydrogen and other explosive gas mixtures vill not be increased (Base 3/4.11.2.6) as a result of the' piping modification. The piping modification vill not affect the restrictions of the gross radioactivity rate of Noble Gases from the Main Condenser as described in (Base 3/4.11.2.7).
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, . PY-CEI/NRR-1141 L Page 12 of 285l SE No.t- 88-407
-Source: Documents- USAR CR 88-227 i'"
. Description of Change
.In accordance with the guidance provided in I.E. Bulletin 88-10 (and as reflected.in PAP-0201, Conduct of Operations)'the operation of the.
Service Air (PSI) System as contaminated must be evaluated relative to-the limits in 10 CFR 20 and 40 CFR 190.
Summary I. No. USAR Section 9.3.1.3 states that-the Service Air System has no safety-related functions and that failure vill not compromise _any.
safety-related-system / component or the ability to conduct a safe .,
reactor shutdown.
The source of contamination was approximately 100 gallons of water from the Condensate Storage and Transfer (P11) System. The source has been isolated and the water drained from the PSI. system. The-postulated radioactive release for this event vould not have
-exceeded any regulatory limit. The sections of piping considered to be potentially contaminated are isolated and steps are being taken to decontaminate them.
i II. No. See Item I above. I III. No. The' operation of the P51 system is not required by Technical Speci fica tions. ;
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, , PY-CEI/NRR-1141 L Page 13 of 285 SE No.: 88-408 Source Document:' DCP 88-0302B, Rev. O Description of Change
-This change removes automatic level control function of the Condenser Air
-Removal (N62) System drain valves IN62-F220A and B by deletion of the nonsafety-related automatic actuation.of solenoid valves 1N62-F050A/B' from the loop seal level switches 1N62-N210A/B and -N225A/B. It installs selector _cvitches in remote panel for manual-remote operation of loop seal solenoid valves.
Summary I. No. The N62 system control logic is not'a part of any: current Perry ;
Safety Analysis reflected in USAR, Chapters 10, 11, or 15.
Therefore, the probability of occurrence or the consequences of an accident / malfunction of safety-related equipment is not increased as evaluated in the USAR.
II. No. The scope of the changes is limited to design modifications described above in Item I. No other system or equipment is affected ;
by this change. All design changes vill remain as nonsafety with !
respect to Perry's design. A different type of accident or malfunction than previously evaluated in the USAR is thus not created.
j III. No. Technical Specifications do not address the design and operability :
requirements of the nonsafety Condenser Air Removal System. The I margin of safety is.thus not reduced.
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Attachm:nt 3 PY-CEI/NRR-ll41 L Page 14 of 285 ;
SE No.: 88-409 Source Document: DCP 87-0783, Rev. 0
-t Description of Change Redesign flex hose mounting to inner Dryvell Personnel Airlock Door (IP53-A3050B).
i Summary I. No. The replacement flex hoses, piping, tubing, and supports are all-ASME Section III, Class 2 components, as are the currently installed items. This DCP relocates a portion of the piping, tubing and supports to accommodate flex hose movement. - It does not change the penetrations or points of intersection currently utilized. This DCP also increases the length of the affected flex hose. This DCP does -l not introduce a new failure mode. Therefore, dryvell bypass leakage is not increased.
These changes are required to preclude flex hose failure experienced in the past. They do not affect dryvell integrity or airlock operation as currently designed. Therefore, the drawing and changes required by this DCP have no impact on the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
II. No. See Item I above.
III. No. This design change does not have an impact on Technical Specification limits or requirements-for the airlock.
7 Attcchm:nt 3 PY-CEI/NRR-1141 L Page 15 of 285 SE No.: 88-410 Source Documents' DCP 88-0075, Rev 0 l Description of Change This design change adds a new interlock between the Residual Heat Removal (RHR-E12) System valves IE12-F006A/B and 1E12-F024A/B which prevents draining the-reactor via the suppression pool return piping while in shutdown cooling. (Mechanical Evaluation) c Summary I. No. This design change adds a new interlock between 1E12-F006A/B and 1E12-F024A/B which prevents draining the reactor via suppression pool return piping vhile in shutdown cooling. The probability of occurrence of an accident (for example, a Loss of Coolant Accident) is reduced by this hardware modification.
Presently, existing plant procedures (S0I-E12) are utilized in preventing the loss of coolant accident via opened F006 and F024.
With this new design, F024 cannot be opened until F006 is fully closed. If the interlock malfunctions (i.e., fall to provide the interlock), reactor water vill flow into the suppression pool via opened F006. The probability of this malfunction is no different than with no interlock- . Operator error must create the initial valve line up which causes the LOCA.
The Loss of Coolant Accident, as stated in Chapter 15 of the USAR, envelopes the LOCA event possibility by operator error along with failure of the designed interlock. The consequences of an accident (10 CFR 100 exposures) are not increased.
Failure of the interlock to perform its function along with operator '
error will drain the reactor vessel while in shutdown cooling.
Reactor water level isolation signal at Level 3 vill close F008 and F009. Thus, the consequences of this failure (reactor water into the suppression pool) vill not cause an Emergency Closed Cooling Systems (ECCS) initiation. '
In the event the interlock fails to allow the opening of the F024 valves during initiation of suppression pool cooling, F024 can be opened manually provided F006 is verified closed.
II. No. This design change does not create a new type of accident or malfunction. This change does not relate to events or disturbances that are considered as potential initiating causes of threats to the fuel and/or the reactor coolant pressure boundary.
III. No. The RHR margins of safety as outlined in Technical Specification L
bases 3.4.9, 3.5, 3.6.3 and 3.9 are not reduced by the addition of interlocks to lE12-F006A/B and 1E12-F024A/B. The Suppression Pool Cooling mode, Shutdown Cooling mode, Lov Pressure Injection mode, and Containment Spray mode vill still meet the original design requirements.
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Page 16 of 285 SE No.: 88-411 Source Document: DCP 88-0075, Rev. O Description of Change This design change adds a new interlock between the Residual Heat Removal (RHR-E12) System valves 1E12-F006A/B and 1E12-F024A/B which prevents draining the reactor via the suppression pool return piping while in shutdown cooling. Further, this change modifies the Safe e Shutdown Capabilities Report (SSCR) with respect to the routing of l circuitry through fire areas covered by deviation requests. (Fire '
Protection Evaluation)
' Summary I. No. This change vill require the routing of circuitry through fire areas covered by deviation requests. However, the intended method of ,
achieving shutdown (that is Method A or Method B) in each of these areas is not altered and the basis for each of the affected deviation requests is not affected. Therefore; the probability of 3 occurrence and the consequences of an accident or malfunction of '
equipment important to safety previously evaluated in the USAR is not increased.
II. No. This change does not affect the Fire Protection Program in a significant way, therefore; the possibility of an accident or -
malfunction of a different type than evaluated in the USAR is not created.
III. No. Only administrative aspects of the Fire Protection. Program and Alternate Shutdown Capability are defined in the Technical Specifications. This change has no impact on the administrative aspects of the Fire Protection Program and does not adversely affect ,
Alternate Shutdown capability. Therefore, the margins of safety as defined in the bases for the portions of the Technical Specification l applicable to fire. protection are not affected.
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.SE No.: 88-412- ,
Source Document: DCP 88-0075, Rev. O Description of Change Safety evaluation from the 16C/ Electrical standpoint to support the DCP described in Safety Evaluation 88-410. (I&C/ Electrical Evaluation)
-- Summa ry I. No. The electrical interlock shown in USAR Figure 7.3-5 (Sheet 3 of 5) for IE12P024B(A), vill automatically prevent both valves from being ;
open at the same time, thus preventing a LOCA into the suppression pool. This action was previously accomplished procedurally. Since this design change reduces dependency on administrative controls, the probability of an accident is decreased.
II. No. The possibility of inadvertent draining the reactor in the Residual Heat Removal (RHR) Shutdown Cooling mode has'been evaluated in USAR Section 5.4 Therefore, no new accident of a different type than previously evaluated is created.
III. No. Technical Specifications are not affected.
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-i Attach::nt 3 PY-CEI/NRR-1141 L Page 18 of.285 SE No.: 88-413 through 88-416 Source Document: See below Description of Change :
Replacement of the existing minimum flov valves listed belov vith a Control Components Incorporated (CCI) " Drag" valve.
SE No. Source Document Description '88-413 DCP 87-0625, Rev.0 Turbine Driven Reactor Feed Pump Minimum Flow Recirculation Valve (IN27-F160A)-
l 88-414 DCP 87-0626, Rev.0 Turbine Driven Reactor Feed Pump Minimum Flow Recirculation Valve (IN27-F160B). t l 88-415 DCP 87-0627, Rev.0 Motor Driven Feed Pump Minimum Flow Recirculation Valve (1N27-F170)-
88-416 DCP 87-0628, Rev.0 Reactor Feed Booster Pump Minimum Flov Valve (lN27-F305)
Summary I. No. USAR Section 10.4.7.2.4 addresses the safety evaluation for the Feedvater (N27) System. The N27 system is nonsafety-related at the subject valves. The safety evaluation in USAR section 10.4.7.2.4 addresses failures within the N27-system, but-does not specifically
-address the failure of these specific valves._ The new feedvater minimum flow valves serve the same function as the old valves and the new valves meets all applicable codes. The new " Drag" valves vill increase the reliability of the minimum flow (recirculation) line by reducing vibration and valve cavitation damage. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not been increased.
II. No. See Item I above.
'II. No. The Technical Specifications (3.6.1.9) address the N27 Feedvater I
Leakage Control System and also addresses (4.6.4.1) Containment Isolation valve operability, IN27-F559A&B. This DCP does not affect the Feedvater Leakage Control or the IN27-F559A&B check valves.
Att:chmant 3 PY-CEI/NRR-1141 L Page 19 of 285 SE No.t 88-417
' Source Documents- USAR CR 88-282 ,
. Description of Change Evaluation'of a USAR change request to Section 17.2.1.3.1 and Figure 13.1-2 to reflect a reorganization of senior management within CEI and Centerior. The President of CEI now reports to a newly named Chief Executive Officer of CEI instead of to the Chairman and Chief Executive Officer of CEI. Cleveland Electric Illuminating Company no longer has a Chairman or Board of Directors.
Summary
.I. No. This is a change in the reporting relationship at CEI's home office.
No changes to the plant or procedures are being made. The Vice President Nuclear Group onsite still reports to the 3 President - CEI at the home office.
.II. No. .This is a change in the reporting relationship at CEI's home office.
III. No. . Technical Specifications 6.2.1.a. and c. are still fully met by this organizational change. No other Technical Specifications or Bases '
are affected by this change.
SE No.: 88-418 Source Document: SXI-0032 Description of Change Performance of Special Test' Instruction (SXI)-0032, Radvaste Programmable Logic Controllers A&B Chemical Vaste Distillate A&B Automatic Program Test.
Summary I. No. The performance of this Special Test does not change any design configuration nor increase the probability of a radioactive liquid release. This is a one time test. Therefore, there is no change in evaluated accidents or malfunctions.
II. No. See Item I above.
III. No. The performance of this test does not affect any liquid radvaste discharges. Any discharges will be controlled under the normal method by SVI-G50-T5266, Liquid Radvaste Release Permit.
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PY-CEI/NRR-1141 L-Page 20 of 285 SE No.: 88-419 Source Document: DCP 86-0561, Rev. 0 Description of Change ,
This design change replaces the existing mechanical, float-type level switches for the containment upper pool. vater level [ Fuel Pool Cooling and-Cleanup (G41) System] and deletes the alarm function of the bistable in the Suppression Pool Make-up G43 System.--The alarm function is not now being used because the switching function of this device cannot reset at the predetermined differential between the high and the lov vater level, thus creating a nuisance alarm.
Summary I. No. The deletion of the alarm function of this bistable does not ,
compromise the system integrity nor the system function since the same alarm function is being performed by the level switches in the G41 system as modified by this DCP. Neither system logics of G41 nor G43 are affected. The volumetric capacity of water in the upper pool, previously established in the USAR, has not been changed. The control system circuit function has'not been changed nor has the previously analyzed safety evaluation been affected. The same design basis has been used. Based on this, the probability of occurrence and/or the consequences of an accident or malfunction of safety-related equipment previously evaluated in the USAR is not increasec.
II.-No. See Item I above.
III. No. The-LC0 for Technical Specification 3/4.9.9 remains unchanged as a result of this DCP. The margins of safety as defined in the Technical Specifications are neither increased nor decreased. This DCP enhances the system operation by providing an' alarm which vill alert the operator during refueling that the Technical Specification limit-in 3/4.9.9 is approaching.
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Page 21 of 285 SE No.: 88-420- i Source Document: LL&JED: 1-88-159 i Description'of Change i Due to a recurring problem with the Motor Feed Pump (MFP) loss of control signal circuitry, the circuit card for IC34-K607C vill be removed by this Lifted Lead and-Jumper / Electrical Device to defeat the freeze circuit until troubleshooting can be performed at the next forced outage.
Summary I. No. The MFP Loss of control signal circuitry failure vould not impact-any other component as described in USAR Chapter 15.
II. No. The MFP normally secured in standby readiness. This LL&JED vill not prevent the use of the MFP in standby readiness.
III. No. The MPP is not required in the bases of the Technical Specifications.
SE No.: 88-422 Source Document PNPP Emergency Plan, Rev. 9 Description of Change Various organizational, administrative and technical changes have been made throughout this revision of the Emergency Plan (OM15A). These changes have been evaluated to ensure that the effectiveness of the PNPP LEmergency Plan has not been reduced, per=10 CFR 50.54(a), and to ensure-that they continue to meet the standards of 10 CFR 50.47(b) and the requirements of Appendix E.
Summary I. No. The Emergency Plan outlines the administrative (emergency preparedness) response to an accident or' equipment malfunction and, therefore, does not affect the probability-of occurrence or the consequences of an accident.
II. No. OM15A does not direct the operation of plant systems or equipment and, therefore, does not create the possibility for an accident or malfunction.
'III. No.- OM15A utilizes existing Technical Specifications and does not control or affect.their revision; therefore, the margin of safety as defined in the-Technical Specifications is not reduced.
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',. , Attcchn:nt 3 PY-CEI/NRR-ll41 L Page 22=of 285 SE No.: 88-423 Source Document: DCP 86-0952, Rev. O Description of Change Install solenoids to interlock the Offgas System charcoal adsorber outlet valves (IN64-F053 A&B) with the bypass and inlet valves. This vill a simplify operation and enhance reliability by automatically closing the
'P053A&B valves whenever the system is in bypass. This vill prevent backflow and moisture intrusion to the charcoal beds and minimize operator action needed during system operation.
Summary I. No. The addition of the two new solenoid valves (located on ,
, valves IN64-F053A&B) serve to close the charcoal bed adsorber outlet valves whenever the charcoal beds are bypassed. This action .(
prevents moisture from entering the beds via the exhaust vent.
These new solenoid valves increase the reliability of the charcoal beds.
If the ncy solenoid valves fail to their de-energized position, the F053A&B valves vill continue to be operated by their current / pneumatic (I/P) controller and the existing solenoid valves.
If both solenoid valves (on each outlet valve) fail, the outlet valves vill go to their. failed open position. The addition of,the solenoid valves vill not affect the safe operation of the charcoal -
bed adsorber outlet valves; therefore, the probability of occurrence of an accident or the malfunction of equipment important to safety vill not be increased.
II. No. The new solenoid valves will not affect the normal operation of-the outlet valves. If the solenoid fails, the valve vill return to its safety position as it would before the change. Therefore, there vill be no new. accident created.
III. No. This change does not affect Technical Specifications. Therefore, the margin of safety will not be reduced.
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1 Page 23 of 285 JSE No.s.88-424 Source Document: DCP 86-0420, Rev. O Description 1of Change Relocate root valve IP45-F643/PI-R709 to upstream of IP45-F523 to provide a more accurate delta P (6p) reading across the High Pressure Core Spray (E22B) Diesel Generator Heat Exchanger.
Summary I. No. This is a system enhancement. Moving the existing test connection '
(root valve IP45-F523/PI-R709) from downstream of-1P45-F523 to upstream of IP45-F523 vill actually improve the op indication across the diesel heat exchanger for corbicula monitoring. This change vill not change the system operability or intended function.
Therefore, the probability or the consequences of an accident or ,
malfunction are not increased.'
II. No. No other design basis parameters are changed by this change.
Therefore, no new accident or. malfunction is created. <
-III. No. Tliis design change does not impact the system' design basis parameters nor the bases for the Technical Specifications. Thus the-margins of safety are not reduced.
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_SE No.: 88-426-Source Document: SCR 1-88-0010T Description of Change This is an evaluation to determine the effect of increasing the calibration factor on the Recirculation Pump discharge line A Loose Parts Monitoring (R63) System channel 7 from 0.7 to 0.8. The purpose of this-change is to ensure that the setpoint is suf ficiently above the normal background noises to minimize spurious alarms, yet lov enough to meet sensitivity-requirements.
Summary I. No. The purpose of this SCR is to raise the Reactor Recirculation Pump discharge line A loose parts channel 7 annunciator setpoint from its present value of 0.35 ft-lb. to 0.4 ft-lb. The setpoint list currently lists a value of 0.5 ft-lb. for the alarm value; however, a 0.7 multiplication calibration factor is used in the field which is not identified in the setpoint list. This SCR will identify a new 0.8 calibration factor in the remarks section which vill make the new setpoint 0.5 foot-pounds times 0.8 which equals 0.4 foot-pounds. This setpoint was chosen to ensure that the setpoint is sufficiently above normal background noises to minimize spurious alarms yet lov enough to meet sensitivity requirements. The 0.4 ft-lb. setting is still conservative to the-0.5 ft-lb. setpoint described in USAR Section 4.4.6.1. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. Since the new setpoint is still within the USAR stated setpoint and no new failure modes have been added, no new accident or malfunction has been introduced outside the scope of the USAR.
III. No. The operability of the loose parts monitoring system is discussed in Section 3.3.7.8 of the Technical Specifications. The above setpoint is not described in the Basis for the LPMS. In addition, the change does not impact the operability of the system. Therefore, the margin of safety has not been reduced.
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Attachment 3 PY-CEI/NRR-1141 L Page 25 of 285 SE No.: 88-427 Source Document: DCP 88-0119, Rev. O Description of Change Replace the existing centrifugal underdrain pumps (P72) with Flygt Vortex Slurry-Pumps in Manholes 3, 5, 7, 9, 10 & 13, and the backup underdrain pumps in Manholes 6 & 11. The' existing underdrain pumps have a tendency to jam, and require frequent. maintenance. DCP 86-093; replaced the pump in Manhole 1 vith a Flygt Vortex Slurry Pump and its performance was reliable and satisfactory. This DCP also replaces a section of flexible hose with rigid pipe (at.the pump discharge) in'each of the above mentioned manholes.
Summary I. No. The failure of all the existing plant underdrain service pumps was considered in the initial design of the Plant Underdrain (P72)
System. It is a Safety Class 3 Seismic Category I Gravity Discharge System, as described in USAR Sections 2.4.13.5.1.e & f. The replacement pumps (Vortex Slurry) have the same pumping capacity as the existing underdrain pumps. The piping modifications of this DCP (replace flexible pump discharge hose with rigid pipe) vill not affect the check valve backflow potential as described in USAR Section 2 .~ 4 .13 . 5 . 5 . The new pumps require a different rated current for power supply than the existing pumps. DCP 88-0119B (electrical) addresses this in its 50.59 Applicability Review. The failure of the new Vortex Slurry pumps vill not impact any safety-related/ safe shutdown systems and the piping modification vill not affect any potential backflows; therefore, no decrease in safety vill be experienced.
II. No. The pump replacements'and piping modifications vill not affect the original function of the Plant Underdrain System, as mentioned in Item I of this evaluation; therefore, the possibility of an accident other than those described in the USAR does not exist.
III. No. The Plant Underdrain System is not addressed in the Technical Specifications; therefore, the margin of safety has not been reduced.
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Attechm2nt 3 PY-CEI/NRR-ll41 L Page 26 of 285 SE No.: 88-428 Source Document: DCP 88-0119B, Rev. 0-Description of Change This design change modifies the terminations of the underdrain pumps (P72) to. allow quick disconnect of the pumps for maintenance.
Also, the fuse and 0/L sizes are increased to accommodate the replacement of the underdrain pumps under DCP 88-0119.
USAR Table 8.3-1 (USAR Diesel Loading Table)'isLrevised to reflect
!. changeout of the 5 HP_underdrain pumps to 9 HP.
F - Summary I. No. Adding 4KV of automatic load to the Division 2 stub bus increases the Division 2 diesel loading by 0.11% vith 38% of available automatic load remaining. Therefore, this slight increase in load does not increase the potential for equipment malfunction important ;
to safety. '
II. No.. Increasing the Division 2 diesel loading by 0.11% does not increase 1
the automatic diesel loading past the rated 7000V, therefore no accident of malfunction of equipment of a different type than those previously evaluated in.the USAR is possible.
III. No. The automatic loading of the Division 2 diesel has 38% of capacity
. remaining. Therefore, the margin of safety as defined in Technical Specifications is not reduced.
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( a' , PY-CEI/NRR-1141 L-Page 27 of 285
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Source Document: DCP 87-0667, Rev. O ;
Description of Change f
This design change provides power to the Machine Shop Turret . !
lathe (L52E0021) and associated lighting and 120 volt receptacles.to the enclosure that houses the lathe at Service Building elevation 620 feet.
Also, required is recoval of door SB-106 security alarm devices. l f
Summary.
l 3 I. No. The removal of n nonsair:ty security alarm device has no ef fect on- l
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any accidents or malfunctions eva!.uated.in'the USAR.
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'[' '11.- No. . Tne removal of'a nonsafety'securtry dmor alarm device vill not
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generat( any new accidents or' malfunctions, because it vas replaced vith a solid vall, therefore, security effectiveness was not Q- '
. effected.
i; III. No.. Door alarms are not addressed in the Technical Specifications and F f have r.o effect on safety. ,
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PY-CEI/NRR-1141 L -
I Page 28 of 285 l t
Source Documents. DCN 2427, key. O q Description of Change !
i Replace IN27-F0602A (1-inch globe valve) with a 1-inch gate valve and !
coupling. '
I i Summary I
' 1. . No. USAR Section 10.4.7.4 addresses the Feedvater (N27) System. This ,
change involves the replacement of IN27-F0602A with a 1-inch gate p[ ,
-valve and coupling.. The N27 system is nonsafet*/-related at the l subjvet vdve (N27, F0602A). The safety evaluation addresses !
fr . f311Gres sithin'the N27 system, but does not specifically addresa ;
- ' ?.he hilure of. the heator drain valve IN27-F06024. The design j intent / control .of the subject valve vill not change due to this MN. ;
Therefore, the probability of occurrence or the consequences of s.n ,
' accident or mall' unction of equipment important to safety previously a evaluated in the USAR has not increased. .
i I I . No. _ ?Se ve.lve replacement vill not change the original function of the D;tedvater System nts mentioned in Item I of this evaluations !
therefore, the probability of an accident or malfunction of a 1
~9 different type '.han any evaluated previously in the USAR does not 1 exist. '
III. No. The Feedvater System still retains its original function with the ;
replacement of the Heater Drain Valve. This portion of the system ;
is not addressed in the Technical Specifications. The modification l vill not affect the operability or the availability of this or other ;
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PY-CEI/NRR-1141 L l ly Page 29 of 285- ,
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P SE No.s. i 88-431 Source Document:
t, LL&JED 1-88-163 FCR 20556 i
Description'of Change i
' Evaluation of an LL&JED to temporarily change the outputs for the RCIC t and RCIC/RHR high steam flow monitoring trip units, 1E31-N690A/B and
-N691A/B, from providing an isolation function to providing an alarm f function. ,
F The GE Design Specification Data Sheet 22A3735AD indicates that these ;
[ trip units should only alarm upon sensing negative steam flov. i i
' Summari-l;
- 1. No. This modification does not irract the steam line. break analysis- l L" described In-Chapter 15 of t!ae usan, The Veneral Electric resporme L . to FCR 30566 .supperts this cor,cinaica - 1
- II. No. This does not create the possibility of an accident or malfunction !
on n dif ferent type thrm any eva'.unted in USAR Chapter 15. i l- .;
III. No. 1This isolation signal is not defined in the Technical '!
Specifications. Furcher, it was not intended to be en isolation j signal per FCR 10556 response and the CE System DesiFn Specification. ;
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-SE No. '88-432 through 88-441 !
Source Document: See belov ,
Description of Change ,
Evaluation of changes made to the pre-fire plan instructions' content for
-the buildings / areas listed belov versus the USAR and Appendix R to 10 CFR 50.
SE No. Source Document Description .88-432 FPI-2AX Auxiliary Building, Unit 1 '88-433 FPI.JDG Diesel Generator Building i 88-434 FPI-0CC Control Complex i
+ ,,08-435 TPI-018 7nterinediate Building -
.x 88-436 FP7-OFH 5'oal Handling Belliing f 1
,,88-437 PI'I-1R9 Reactor Btdiding l 03-436 FPI-0EV Emergency Service Water Pamphouse 88-439 FPI-0YD Diesel Generator Fuel Oil Storage ,
Tanks ;88-440 FPI-IST Ster,m Tunnel 88-441 FPI-1 CST Condensate Storage Tanks, !
. Unit 1 and 2 t Summary i I. No. The changes being made either meet or exceed the requirements in ;)
Appendix 9A of the USAR and in Appendix R of 10 CFR 50 for >
pre-planned strategies for fighting fires in specific plant areas. >
As a result, the probability of an occurrence or the consequences of !
an accident or malfunction previously identified is not increased. 1
-II. No. These changes do not create the possibility for an accident or ;
malfunction of a different type than any previously. identified in !
the USAR. The changes being made are improvements and are -!
consistent with the requirements of Appendix 9A of the USAR and also '
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PY-CEI/NRR-1141 L l F Page 31 of 285 .
SE No.88-432 through 88-441 (Continued) ,
, Summary l
i III. No. The changes being made vill result in an improved. method of o providing pre-planned fire fighting strategies over the current u
E pre-fire. plans that are being used. Both the present and revised ;
methods meet the requirements of Appendix 9A of the USAR and L Appendix R of 10 CFR 50. Fire protection program is referenced in 1 Section 6.5.1.60 6.5.2.8 and 6.8.1 of the Technical Specifications. ;
This evaluation has determined that all changes made to these" ;
procedures and the Fire Protection Program by these procedure i changes are consistent with these sections of the Technical ;
- jlf Specifications, j
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Attcchment 3 N- PY-CEI/NRR-1141 L Page 32 of 285 SE No.: 88-442 '88-442, Rev.-1 Source Document: DCP 88-0051, Rev. 0
- DCP_88-0051, Rev. 2 Description of Change f
Add 21 branch isolation ball valves to the Service Air (PSI) System to facilitate maintenance of the system. This will make it possible to i repair a portion of the system without shutting down the entire system.
Revision 1 to this safety evaluation corrects a duplication of the Master Parts List (MPL) identifying numbers between two DCPs and the P&ID vas. ;
revised to reflect the correct HPL numbers. l Summar_y c y
- 1. No. The addition of these air isolation volves does not increase the O
probability of occurrence of an accident. Addition of these valves allovs the isolation af branches of the Service Air System to facilitate maintenance. This change inc*: cases the safety of tne i plant by allowing the-isolation of f aulty brarahes without removing ,
the entire syttem frora service. The consequences of an acci. lent are reduced by building flesibility into the system.
/.. malfunction of any of the new isolation valves-voeld not affect any safety-related components / equipment as the Service Air System it :
not relied on to support the operation of safety-related compontnts/ equipment. .
Revision 1 to this safety evaluation changing USAR Figure 9.3-29 to show all the new ball valves and changing 3 HPL numbers does not alter the previously evaluated safety considerations for this change.
II. No.- The Service Air CJstribution System, af fected by this DCP, does not have a safety-related function. Per USAR Section 9.3.1.3: failure ,
of this system vill not compromise any safety-related system or component and vill not prevent safe reactor shutdown. Therefore, the function of safety-related components / equipment can not be i affected by this DCP. The system does have safety-related containment and dryvell isolation equipment, but this DCP does not alter or affect any of this equipment. Therefore, the creation of a previously unidentified safety-related accident or malfunction is not possible.
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?t Attcchment 3 PY-CEI/NRR-1141 L Page 33 of 285 i SE No.: 88-442 p5 88-442, Rev. 1 (Continued) '
n i lE Summary e
t III. No. The margin of safety designed into the components / equipment I I
identified in the Technical: Specifications is not decreased as this ;
DCP increases the reliability of the Service Air System. This DCP does not change any information in the Technical Specifications or-i.
t, alter previously described requirements.
e Revision 1 to this safety evaluation changing USAR Figure 9.3-29 to i show all the-new ball valves and changing 3 MPL numbers does not -!
affect the Technical Specifications in any way. This portion of the !
service air system is not in the Technical Specifications. !
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- , Source Document
- DCP 88-0190, Rev. O j Description of Change j Add a vjre mesh enclosure to the Auxiliary Building on Elevation 574 to provide a secure storage area for the Unit 1 calibration standards used j L
in Inservice Inspections. '
Summary I. No. This change adds a vire mesh enclosure to the Auxiliary Building, ,
elevation 574 (Fire Area 1AB-Ig) to be used as a noncombustible l storage area for the Unit I calibration standards. This aren vill also contain a metallic desk, plastic storage bins, vood laydown t blocks and papervork associated with the operation of the facility.
l The total quantity of combustibles installed in this enclosipo vill ;
) be less than 100 lbs. The area affteted by this change (Fire -
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Area 1 Ab-lg) doiss cent?.in rW.undant safe shutdown cirecits/ >
equip 9ent which are not protated in accordance with Appendix ? Jn that automatic suppttssion is not provided. A deviat. ion request hr.s been filtd and accepted by the NRC justifying th2 Inr.k of autor;atic sur.p m c.len be. sed on the lov hazard in this area, physical '
separati m of redundant safe shutdown circuits and lack of GFposure hazard in this ared. .
The b.esis of this exemption request is not changed by this DCP '
considering that this storage sroa is separated from Appendix R equipment and cables by a distance in excess of 20 feet and it ,
centalus less than 100 pounds of ordinary combustibles. A fire '
involving the storage area vould not be of sufficient magnitude to adversely affect Appendix R equipment / cables located in this fire area, therefore, it need not be considered as an exposure hazard.
Therefore, the probability of occurrence of an accident or ,
malfunction of equipment important to safety previously evaluated in '
the USAR is not increased and the consequence of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. This change does not adversely affect the Fire Protection Program, therefore, the possibility of an accident or malfunction or a different type than evaluated in the USAR is not created.
III. No. The only aspect of the Fire Protection Program addressed in the bases section of the Technical Specifications is alternate shutdown capability. This change does not adversely affect alternate shutdown capability, therefore, the margin for safety defined in the bases section of the Technical Specifications is not decreased.
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t Attcch::nt 3 PY-CEI/NRR-ll41 L o Page 35 of 285 SE No.: 88-444 through 88-447 Source Document: See belov Description of C5a m These des;gn changes to the reactor vessel level instrument reference legs listed belov, replace the existing temporary N14 nozzle deflectors with N14 nozzle deflector shields and also replace some of the associated reference leg piping. These nozzle deflector shields prevent water from entering the reference leg during RCIC head injection, causing water
- level measurement problems.
F SE No. Source Document Description 88-444 DCP 87-0385, Rev.0 Reference leg D 88-445 DCP B7-0584, Rev.0 Reference Jeg C q
98-446 DCr B?-0363, hev.0 Reference Icg 8 - 88-447 DCP 87-0382, Pov.0 Fetereine leg .\
Summary I. No. Instnation of the deflector shields vill be performed in acccrdance with tr.e r2quirrments of ASME Section ITI; Subsection NB, ASHE Section XI. The material used in the installat'lon meets the requirements set forth in ASHE Section II, Part A. The velding of the deflector vill be performed by velders qualified to ASHE Section XI. Inspection of the velding vill be in accordance with ASHE Section V. The installation of the deflector shields serves the same purpose as the presently installed nozzle insert shields which have proved effective in preventing the reactor vessel level anomalies experienced during RCIC injection. The new design vill be tested to prove its effectiveness. The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. See Item I above. III. No. Since operation of the system has not changed and the installation of the deflector shields vill assure proper operation of the reactor ' vessel level instrumentation, the margin of safety as defined in the Technical Specifications is not reduced, l l l
b Attechment 3 PY-CEI/NRR-1141 L l Page 36 of 285 SE No.: 88-448-Source Document DCN 2464, Rev. 0 Description of Change i This drawing change involves an editorial change to P&ID 302-081. As presently indicated on the line connecting flags for the Reactor Feed Booster Pump (RFBP) logic, connector flag "B" has two different functions, RFBP auto start on low net positive suction head (NPSH) and i [ RFBP Suction Strainer delta P. These functions will be split out as "B" i [ and "B2" for clarity. i. Summary ! I. No. This. drawing change is an editorial change to P&ID 302-081 to clarify the line connecting flags. The connecting flag denoted as I o connector "B" is used for two different logic functions, RFBP e suction r. trainer delta P and RFDP neto start on Icv NPSH. The line , conne:: ting flag for RFEP. suction strainer delte P will be changed to _
- ronnector "B2". TSis vill t.hangc '.he P6ID (302-001) shown as
", Figure 10.1-3. Sheet 1 of 2 in the t:SAR. This chaaga is required : for clarification. Thic DCN does not change th7 des. tan intent / function of the control logic for the RFBP's. Thartfore the i raobability of occurrence or the consequemes of em ucident or malfunction of equipment importt.nt to sr.fety previour.ly cvaluated in 4 ; the USAE hrs not bet-n increased. -i II. No. See Iten I ahave, this is an editorial change to the PEID which is i in the USAR. This is not a design change. Therefore, there is no i possibility for a different accident or. malfunction from the one E previously evaluated. III. No. The RFBP's are not addressed in the Technical Specifications. . l E I
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i Attcch::nt 3 PY-CEI/NRR-1141 L ! Page 37 of 285 ;
.SE No.:
88-449 Source Document: USAR CR 88-283 CR 86-703 Description of Change i i Evaluation of a clarification to the USAR Section 6.7.3.1 regarding ; physical separation of MSIV-Leakage Control (E32) System (LCS) inboard and outboard system components. Summary '
!I I. No. In the SER (Section 6.7), the NRC stated that "the components of each subsystem are protected by separation and barrier against ,
internally generated missiles, externally generated missiles and dynamic effects associated with pipe breaks, so that their function ' is not impaired under postulated LOCA conditions. Thus the re>quirements of GDC-4, Environmental and Missile Design Bases, and the guideliner of Regulatory Guide 1.96, Positions C.2 and C.4, are
- autisfied." .
The Perry desit,a does n:t include barriers and the beparationa, however, Perry OSAR statemente relating to the above t.rlteria ccncerning inboard / outboard sys'vm sepuration nust be clarified to -l reflect existing design, sinco some redanJant ceripouents are t actually only a small distance apart in cartain cases. Th2s Anfr:ty l Evaluttion shavs hov Porr; rtill meat.s GDO-4 cnd Regulatoty ? Guide 1.9(, Pos ?. t iott.= C 7 and C.4 vitn eaupect to externally - ge.terated misciles, intornally generated miesflen and the dynamic ' eftects associated vith p1rm breahs for this system vith existing design. An anulysis of these wffects follows: Externally Generated M19siles The entire system is located within the Auxiliary Building-and the l building structure provides protection from externally generated ; missiles. Pipe Break Dynamic Effects Vithin the room with the adjacent blowers, no jet impingement can affect both subsystems (Steam Tunnel), dynamic jet loads are part of the original design loading combinations (see DSP-E32-1-4549). Internally Generated Missiles No other equipment in the vicinity of the adjacent blowers capable of generating a missile of enough magnitude to damage both trains has been identified. Therefore, the possibility of a blower missile ~ vas investigated as the only remaining threat, with the following results:
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Attcchment 3 . ! PY-CEI/NRR-1141 L I Page 38 of'285 i ) 1 SE No.: 88-449 (Continued) { l Summary Discussions with GE's vendor ~ "Siemens" indica' e that no burst - failure or penetration of the cast. iron housiags have occurred with , this model blower which has had_a long servlee history. This is ; despite the fact that the blowers have haen' applied on variable 'l speed drives with speeds to estimated 5,000 rpm. The Perry 1 application is motor driven with no potential for overspeed. !
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.The manufacturer indicated that known failures involve entry of a )
foreign object into the blover which may jam the impeller, break the impeller blades or both. This generally results in seizure of the ,
; blower and frictional melting and weld-like adhesion of the aluminum I?
impeller surfaces in contact with the blover hcusing. This thermal dissipation of energy and stalling of the motor drive does not' , produce missiles. To confirm the vendor information, an IMPO-NPRDS survey was ' performet. Tine survey found that of 18 identical or similar Siemons
~ blowers 'u vcrious pla.its surveyed, only one f ailure occurred. This :t vas due to bearing fallare within the cacing ne described above (S9aquehanna Unit 2). 3 NED Calculation E32-9 concluded t.he blovers do not pose a missile hazard to redundant trains. Therefore, based upon this analysir., )
Perry meets G00 4 and Regulatory Guide ?. 96, Position C.2 and C.4 vith-respect to the MSIV-LCS without the barriers and separation assumed by NRC in the SFR. II. No. See Item I above. III. No. The Technical Specification margins of safety are maintained. 0' 1. l
K 4 Atttchment 3- ! PY-CEI/NRP.-1141 L Page 39 of 285 SE'No.: 88-450 t L Source Documents- DCP 88-0252, key. 0 i Description of Change i This. design change constructs an insulated metal enclosure around the ; existing Main Steam Valves (MSIV) removal steel to provide a closed < t; environment for working on the MSIV's. ; Summary : I. No. -The changes to the USAR required by the implementation of this [
' design change are limited to the updating of five of the plant-layout figures to reflect the existence of this new room.. ,
h L The design of this new enclosure includes consideration'of missile and seismic concerns, as well as the integrity of the existing MSIV' ! removal and Auxiliary Building structures. Based on this review, y the existing structures are adequate for the new loads imposed on ' ' .them by the new enclosure, and therefore the' probability or consequencas of an accident previously ovaluated are not increased. . II. No. For the same reasona outlined in Item 1 above, the possibili.ty of an accident or malfunction of a difftre:it type than any evaluated -t previously'ir not created, i t i
- III . ' No . - This DCP simply adds cn enclosure around .the existing MSIV rernval- l 1
steel to-eteate a favorable environment for maintenance personnel. .l As such, this change does nat affec*. any of the plant- operability ! limito described in the Tetanical Specifications and therefore no ' Technical Specification margins of. safety are reduced. 4 l
?
e 1
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i t Attcchment 3 PY-CEI/NRR-1141 L Page 40 of 285 SE No.: 88-451 Source Document: USAR CR 88-028 i Description of Change Evaluation of a USAR change request to Section 3.6.2.1.5.a.5 to indicate that design changes made during operations vill no longer postulate arbitrary intermediate pipe breaks. Also, previously postulated arbitrary intermediate pipe breaks may be eliminated if they result in improved plant maintenance capability. These changes are made to correspond to NRC Generic Letter 87-11, which describes a revision to Branch Technical Position MEB 3-1 of the Standard Reviev Plan (NUREG-0800) Section 3.6.2, which eliminated this requirement. Summary I. No. Arbitrary break locations were originally required by the NRC in SRP 3.6.2 if at least two locations between terminal end points did not exceed the break criteria. The NRC subsequently realized that i these arbitrary break locations could actually increase the probability of an accident by restraining thermal expansion or making velds inaccessible to inservice inspections due to interference with rupture restraints and/cr source shields. As a result, they revised SRP 3.6.2 to no longer require postulation of arbitrary intermedicte breaks (per Generic Letter 87-11). The consc(vences of postulating arbitr6ry breaks manifest themselves ' in the form of pipo whip, jet incingemen', environocntal snd flooding effects on equipment importr.nt to safety. .te addition, - lors of reactor coolant and offsite dose cansequecces may alvo acsolt. LOCA and offsite dose consequences of these breaks remain boundea by design basis events nnalyzed in Chapter 15. The other effects are presently prevented from affecting equipment important to safety by restraints, shields, physical separation and flooding protection (e.g., curbs;. These arbitrary breaks had no logical basis for existence in the original analysis. Therefore, their
- elimination does not pose any additional risk. Neither the consequences of an accident, nor the probability of occurrence has been increased.
II. No. 'This change does not create an accident of a different type. The type of accident (pipe breaks) is thoroughly evaluated in the FSAR. In addition, since the arbitrary breaks are no longer required to be postulated, there can be no effects on nearby equipment important to safety. Therefore, no new type of equipment malfunction is involved. III. No. The margin of safety of piping is determined by rules of ASME III. These rules are unchanged by elimination of the arbitrary breaks.
c , [.', at j Attcchment 3- ! PY-CEI/NRR-1141 L I Page 41 of 285 ' SE No.: 88-452 l
-Source Document: USAR CR 88-285 l
p Description of Change L . I i Evaluation of a USAR change request to Appendix 1B, Item 15 to indicate ; that the NRC has reviewed and approved the PNPP Leak Reduction Program l and fulfilled the comm*tment described in License Commitment.15 in the : USAR. i l Summary ' I. No. This change documents closure of a License Condition. No change to the plant or procedures is being made.
'II. No. This change documents closure of a License Condition. No change to '
the plant procedures or testing is being made. >
'I III. No. The leakage. identified by this program has been included in !
secondary containment bypass leakage'and is thus accounted for in total leakage rate for systems outside containment. This change l documents NRC's approval of the Petry Leakage Reduction Program. Technical Spccification 6.8.3.a is satisfied. ! I
.i , I i
4 b i e 4 L a m - e , - .,-,
if [t Attechment 3 , PY-CEI/NRR-ll41 L Page 42 of 285 i I- SE No.: 88-453
. Source Document: DCP 88-0103, Rev. O I o Description of Change
(
! Replace the keylock switch 1E32-S4 for the MSIV-LCS with a spring return i f- to normal hand operated switch. Switches lE32-SIA, SIE, S1J and SIN vere [
previously replaced under an FDDR. This design change corrects a Human ! l . Engineering Deficiency Report (HED-370) and is a licensing. commitment. t
! Summary I. No. As the electrical portion of the new switch.(contact block -
assemblies) is identical to the old switch part, and the operator is ' manufactured to the same quality requirements as the old switch, the '
!. probability of malfunction is not increased.
As system operation is not changed by this DCP, the probability or. I consequences of an accident have not: increased following , implementation of this change. ; The probability of accidental initiation of this system is reduced by performance of. this change. According to Human Engineering Deficiency (HED) report HED-370 Eev. 2, leaving the key in a keylock l switch creates possibility of switch activation due to bumping the switch. Additionally, the HED report states that keylocked switches should be used for " bypass restraint ar.d test functions." As these svitches do not meet these criteria, keylock switches.in these - upplications could cause operator confusion. II. No. -Since the electrical functicn of the switch has not been altered or [ changed by this modifice. tion, no pov or dirferent malfunction or accident is created. Systein onoration is not changed by thic DCP. ' III. No. The switch modificatior does not change the function or intent of the original design. Therefore the margin of safety is not reduced. j System operation is not changed by this DCP. l l 6 l l i l-t
c o [c f'1 - Attcchment 3 L PY-CEI/NRR-1141 L ! Page 43.of 285 r SE No.: 88-454 , Source Document PAP-0204, Rev. 4, TCN 1 USAR CR 88-300 Description of Change Evaluation of a USAR change request to Section 5.2.5.1.1 and a temporary change notice to PAP-0204 Housekeeping / Cleanliness Control Program, to ' delete the requirement for material accountability in the dryvell during maintenance. Potential for design system blockage is adequately L addressed by the dryvell closeout required per 101-1 and 2, and the [ present method of sump level and rate measurement. Summary i L L I. No. This change does not increase the probability of occurrence or the consequences of an accident or malfunction. The FSAR Q&R ASB-3 which addresses section 5.2.5 - Reactor Coolant Pressure Boundary Leak Detection is met by the following three items. . First, leakage into the surps is monitored by level transmitters 1E31-N093 and 1E31-N094. The level signal is also used to indicate rump fill rates by use of dynamic compensators which operate in the rate mode to indica'e rate of change of the level indication. This method of level and rato measurement results in continuous level / rate observation. Secondly, the dryvall sumps are inspected af te'. ermplatf or, of all work in the dryvell prior to dryvell cloteout per Attechment 2 of 101.1 or 2. The sumps are inspected for water havel, debris, oil and pi'oper operation of - : equiparat. Thirdly, discharge pressures on all four drvvell sump pumpt are monitored by pump discharge pressure switches 1G61-il280 and IGGl-N29r. Theen preraure svitches' trip the purps which indiccte bioekage in the discharge piping. These three items ensure that the sumps are not blocked and that ; indication of sump operation is maintained. Therefore, it is not necessary to maintain strict material control practices in Plant Condition 4 or 5 to minimize the potential for drain system blockage. II. No. This change does not present a possibility for an accident or malfunction of a different type than any evaluated in the USAR. III. No. This change does not effect any bases for any Technical Specification. s
F- ;B V p Attcch: nt 3 PY-CEI/NRR-1141 L i Page 44 of 285 SE No.: 88-455 Source Document DCN 2313 Rev. 0 l Description of Change
, This drawing change reflects the added diverter piping to air vent valve P 1N71-F0527B in the circulating Vater (N71) System.
-Summary L 1. No. USAR Section 10.4.5.3 (Safety Evaluation for the Circulating Water System) does not address the system's piping auto air vents or their g
diverter piping. The auto air vents are used during system fill and venting. The diverter piping addressed in this safety evaluation is located in the Heater Bay. The diverter piping is-used to route water, if any, which may be blown out, away from equipment. This piping vill eliminate the possibility of electrical equipment damage which is located within the Heater Bay. The diverter piping is designed / installed to B 31.1 code. The diverter piping does not change the design intent of the circulating Vater System. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not increated.- II. No. See Item I above. III. No. The circulating Water System is not addressed in the Technical Specifications. A i L !
r Atttchment 3 p PY-CEI/NRR-1141 L i Jage 45 of 285 i SE No.:- 88-456 Souice Document: DCP 88-0117, Rev. O p Description of' Change I Install carpeting in the operations Support Center (OSC) to reduce noise i during Emergency Plan Drills. Summary I. No. This change increases the combustible loading in Fire Zone CC-2a by y a small amount. Therefore, this area vill remain a lov hazard area. [ Considering the fixed suppression capabilities in the vicinity of the carpet added by this change, manual fire fighting capabilities p ' b and the overall low hazard in this area, the fire hazard analysis i vill not be adversely affected. ' e j. Therefore .the probability of occurrence or the consequences of an c, L accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. This change does not affect the Fire Protection Program in a ; significant way therefore, the possibility of an accident or ' malfunction of a different type than evaluated in the USAR is not ' created. , III. No. No, only administrative aspects of the Fire Protection Program and , the remote shutdown system are addressed in the Technical .! Specifications. This change is consistent with these sections of the Technical Specifications. - SE No.s. 68-457 -' Source Documens: DCN 2267, Rev. 0 [ Description of Change Editorial correction to drawing D-352-241 (USAR Figure 9.3-1) concerning '
~ delta gate valves 2P52-F524A/B and 2P52-F526. The note on the drawing presently identifies them as ball and plug valves. These valves are
.actually gate valves.
Summary I. No. Correction of this editorial labeling does not create a potential for an accident. II. No. The valves were and remain gate valves. III. No. The Technical Specifications do not address these valves, b 5 I t
Attcchment 3 [- PY-CEI/NRR-1141 L Page 46 of 285 SE No.: 88-458, Rev. I through 88-460, Rev. I ! Source Document: See belov Description of Change _ Evaluation of the removal of the open side torque switch from various Motor Operated Valves (MOV) and removal of the limit switch compartment i space heaters if found installed. ' SE No. Source Document 88-458, Rev.1 DCP 87-0687, Rev.0 88-459, Rev.1 DCP 87-0687A,-Rev.0 88-460, Rev.1 DCP 87-06878, Rev.0 Summary I. No. SOER 86-02 " Inaccurate Closed Position Indication on Motor Operated Valves" addresses the offects on Rotor 2 of a two train limit switch. When Rotor 2 is adjusted to set the opening torque switch bypass (Limit Switch contact 5) to properly operate an MOV during its unscating, the Control Room close position indication (Limit , Svitch contact 7) light is improperly adjusted. This is due to the design of the limit switch (LS) rotor where contacts LS 5, 6, 7 and 8 are all located on the same rotor, therefore, by adjusting one contact all other contacts are affected. To resolve the above concern, EDCR/DCP 87-0687 vas initiated to remove the open side torque switch and torque svita.h bypass contact , fro'n the MGV's control circuitry. By eliminating the torque switch bypr.ss rtquitement (ecatset 1.5 5), proper Control Room indication can then be achieved by iteeing Limit Switch Rot.or 2 which in tu'rn may be adjusted to properly indicate the valve's full close position (LS contact 7). ' Thr, disadvantage involved with the removal of the open side torque switch from the MOV's circuitry is that there vill be less protection provided to the motor operator. Should a enlfunction occur during the opening cycle of the valvo, the potential for motor burnout vill be increased. Based on all the following concepts, however, this does not cause any increased safety concerns.
- 1. Present design for those valves that have a safety function to open and close, already adopt the circuit configuration of not utilizing the opening side torque switch.
n , o. ; Attcchment 3 i PY-CEI/NRR-Il41 L l
... Page 47 of 285 ;
SE. No.: 88-458 through 88-460 (Continued) , Summary f t 2.' If a malfunction occurs, it vill most likely happen during the unseating of the valve. Under present design and NRC. ' requirement of IE Bulletin 85-03, the torque switch is bypassed +
! until after unseating. l t'
- 3. . If mechanical binding occurs while cycling open, it vill most .
likely prevent closure regardless of the condition of-the i t- motor. V .
- 4. Preventative maintenance program and testing activities are in-p place to identify and prevent arbitrary mechanical binding.. -
5.= Fuse blow out vill most likely occur prior to motor burn outt d therefore, the motor vill be protected and the MOV may be t cycled after replacement of the fuses. II. No; See Item I above.
- i III. No. This DCP vill not change the Technical Specifications. Removal of -
t' the open torque switch will not affect the closing isolation time requirement as described in Table 3.6.4-1. , i i F { i I-l
~
y, e Att chment 3 PY-CEI/NRR-ll41 L Page 48 of 285 SE No.: 88-462
= Source Document: LL&JED l-88-180 L
Description of Change Calculations have been performed in accordance with Regulatory Guide 1.78 which indicate that Ethylene Oxide need no longer be considered a factor vhen evaluating Control Room habitability. This lifted lead and jumper modifies the Ethylene Oxide Monitors to provide an alarm only, instead of h actuating the Control Room HVAC (M25/26) System in the Emergency Recirculation mode. Summary I. No. In accordance with Regulatory Guide 1.78 analyses (CL-M26-002) have been performed that show that Ethylene oxide need not be considered when evaluating the Control Room habitability. These analyses show that the weight _of Ethylene oxide released during a postulated accident at the closest railroad site, 3 miles from the~ plant, is lower than the criteria presented in Regulatory Guide 1.78, Table C-2, and does not become a hazard to the Control Room habitability
- through: the plant's control Room IIVAC System. The release quantity-used for this calculation / analysis is given in Section 2.2 of the USAR. Since there is no increase in the previously evaluated probability for this postulated accident, the automatic control for the actuation of Control Room HVAC in recirculating mode may be removed.
II. No. Toxic gas vill still be monitored and operators vill be warned of its presence. III.-No. The margin of safety is not reduced because toxic Eas is still monitored, but it is not required to actuate an ESF per Regulatory Guide 1.78. l t. L
Attachment 3 PY-CEI/NRR-1141 L Page 49 of 285 SE. No.: 88-463 Source' Documents PAP 1918, Rev. 1, TCN 1 Description of Change Clarification to PAP 1918, Fire Drills, which deletes a confusing paragraph which intended to describe that Fire Brigade members participating in a drill, should not be the same perJons actually assigned that responsibility on that shift. Summary I. No. The deletion of section 6.3.2, page 6 of PAP-1918 vill prevent a misconception that the on-shift Fire Brigade is not being trained as a unit. The requirements vill still meet or exceed Appendix 9A of the USAR and Appendix R of 10 CFR 50. As a result, the probability of an occurrence or the consequences of an accident or malfunction previously identified is not increased. II. No. These changes do not create the possibility for an accident or malfunction of a different type than already previously identified in the USAR. The change being made is consistent with the requirements in Appendix R of 10 CFR 50. III. No. The change being made vill result in a clear understanding of the on-shift Fire Brigade being trained as a unit. The margin of safety as defined in the bases for any Technical Specification is not reduced. I l l r l
Attcchment 3 PY-CEI/NRR-1141 L Page 50 of 285 SE No.: 88-464 Source Document: FCR 10764 Description of Change
' Evaluation of a Field Change Request (FCR) to determine the effect of ,
suppression pool temperature falling below the USAR specified limit of 60'F. Summary I. No. A review of the plant design basis, for primary design considerations relative to the low pool temperature limit was completed. There were two areas where lov suppression pool temperatures vere utilized as primary design input. Each are reviewed below: Containment Negative Pressure Design Limit /CVR Valve Sizing 1 A bounding calculation (Calculation T23-01, Rev. 0) was completed to verify that under the plant operating conditions which existed while at 58'F pool temperature, containment negative pressure would not have been violated given an inadvertent initiation of containment spray. Results indicated there was significant margin between the design limit and the calculated vorst case value. This margin was ' greater than the margin calculated under design basis conditions. Based on this bounding calculation, the design basis calculation for containment vacuum relief valve sizing remains limiting. The 58'F pool temperature has no effect on vacuum relief valve sizing. ; i LSMT - Containment Vessel
- The Permissible Lowest Service Meta) Temperature (PLSMT) for the conteinment vessel as stated in USAh Section 3.8.2.6.1.2.a.3 is-50'F. Based Un the lovest identified bulk pool temperature of
- 58'F, the design limit for the containment vessel was not violated. I No design concern exists.
Sevoral other secondary calculations' utilized the 60'F pool limit as design input. They are discussed below, no other design calculations or analyses have been identified which utilize the lov suppression pool temperature limit. 2.4.4.4.1 R/0 - Foundation Mat Thermal Gradients 2.4.4.5 R/0 - Dryvell Vall
Atttchment 3 PY-CEI/NRR-1141 L Page 51 of 285 [e: , t SE No.: 88-464 (Continued) p g Summary r [" These calculations evaluated the thermal gradient effects on structures / components in the pool atea under LOCA conditions. Based
' on the plant conditions at the time of the 58' bulk pool.
temperatures, the conditions for these gradients did not exist. No design limits vere or could have been exceeded. J Calculation 5.13.4 Rev. 2 - Reactor Heat Exchanger Relief Valve. Discharge Line Water Column Rise, also utilized suppression pool lov
, temperature limit as design input to identify vorst case conditions.
F _ Based on plant operating conditions at the time of the lov bulk pool-temperatures, the condition for RHR relief valve discharge did not exist. No design limit was or could have been exceeded. Based on the above review, at no time was the plant in an unanalyzed condition.- Therefore, there has been no increase in the probability -i
- of occurrence or consequences of an accident or malfunction of equipment important to safety.
II. No. There is no potential for any new type of accident or malfunction. III.-No. The margin of safety for this condition is greater than under design basis conditions. ' k', v 7 k
p l' L Attcchment 3 t PY-CEI/NRR-1141 L p Page 52 of 285 SE No.: 88-465 Source Document: DCP 86-0949, Rev.0 Description of Change To enhance the operability of the Redundant Reactivity Control (C22) System (RRCS) Anticipated Transient Vithout Scram (ATVS) vessel level transmitters by eliminating the possibility of a scram due to spurious level perturbations. The (4) ATVS transmitters have been physically ! separated and assigned to separate panels and reference columns. Scram 1-86-3 resulted from the inadvertent tripping of two ATVS level transmitters by a spurious Icvel perturbation imposed on a common reference column. Summary I. No. The Anticipated Transient Vithout Scram (ATVS) transmitters have been physically separated and assigned their own panel and reference. % column as follows. Transmitter (MPL) Panel No. Ref. Column 1B21N402A 1H22P0004C IB21D0004A IB21N402B 1H22P0027 IB21N0004B 1B21N402E 1H22P0005/P0061 1D21D0004C 1B21N402F 1H22P0026 IB21D0004D This will eliminate the possibility of a scram due to sputicus level perturbations. Since this design change addresver only the physical arrangement of the ATVS transmitters (i.e., transmittet lucations, tubing, fittings, valves, etc.), the transmitter actuation log?c and the necessary interf ace logic (or those syJteme required to perform specific functions in response to ar ATVS event vill remtir as designed. General Electric reviewed this change from a safety and rystem reliability standpoint (Reference FDDR KL1-6572), linder justification of disposition decision (safety, neliability), GE also cencluded that. .. " Safety is net degraded by this changc". This change has beon reviewed against the following sections of the USAR:
- 1. 7.6.1.12 Anticipated transient without Scram (ATVS) - Instrumentation & Controls.
- 2. 15.6.2 Instrument line pipe break.
- 3. Appendix 15C Anticipated transients without Scram (ATVS)
- 4. Regulatory Guide Physical independence of Electrical 1.75 systems.
Attcchment 3 PY-CEI/NRR-ll41 L ' Page 53 of 285 , SE No.: 88-465 (Continued) l Summary Based on the above review, there is no evidence which supports or concludes that by physically separating the four ATVS transmitters ! existing equipment or systems vill be adversely affected. Therefore the probability of occurrence or the consequences of an t accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. Except for the physical location of the ATVS transmitters, the operhtion of the system vill remain exactly as previously designed. The installation (i.e., mounting, tube routing, materials, etc.), shall meet or exceed the current installation requirements for safety-related installations. Therefore, the possibility for an accident or malfunction of a different type than any evaluated previously in the USAR vill not be created. III. No. This design change addresses the physical arrangement of the RRCS vessel level transmitters (ATVS) only and does not affect any operational or surveillance requirements defined under Section 3/4.3.4 of the Technical Specifications. Therefore, the margin of safety as defined in the basis for any Technical Specifications is not reduced. E f Y
{ j i Att:chznt 3- l l PY-CEI/NRR-1141 L i t Page 54 of 285 l SE No.: 88-466 Source Document: DCP 87-0276C, Rev. O Description of Change This design change modifies the Division 1 Diesel Generator Pneumatic control System. [' The scope of the change is to minimize the number of active pneumatic components. Nine solenoid valves, three pneumatic valves, and twenty-six I pressure switches were removed and replaced by eleven electro-Dechanical relays and manual switch contacts.
-Summary i
- 1. No. The pneumatic ccmponents were replaced by electro-mechanical relays and manual switch contacts. The replacement components maintain the same redundancy and function of the original design with significantly fever active parts. Further, the new parts are of equivalent qualification and reliability as the original designed components. Since component reliability and qualification of the parts installed by this design change are shown to be equivalent to l or better than the original design, it is concluded that the i performance of the Division 1 Diesel-Generator Control System is enhanced as the result of this design change.
Therefore the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not inu uased. II. No. All of the en gonents installed as tSt result of this change meet all of the original equipment qualifit.ation requirements as stated in Itern I above. The physical installation of thcae components and their associated tubing and viring is in accordance with the original installation requirements. This change creates no
- reductions of redundancy or component performa^nce levels compared to the original desigr,. Yttertfere, no new potential for equipment malfrnetions is intreduced by this design change.
I.TI. No. The margiu of saf ety as cefined in the bases of Tet huical s Specification Section 3/4-8 refers to the reliability of the onsite pover supplies. As str.ted in Items I and II above, reliability of the Division 1 Diesel Generator is not compromised by this design change. flence, the margins of safety described in the Technical Specification bases is not affected. i i
p Attachm:nt 3 PY-CEI/NRR-ll41 L Page 55 of 285 ' SE No.: 88-467 Source Documents- DCP 88-072, Rev. O Description of Change r This evaluation analyzes modification made to the Division III High Pressure Core Spray (HPCS) Diesel Generator (DG) start circuit. Summary I. No. This change makes two major modifications to the HPCS DG start circuit. First, a parallel set of manual engine control switch contacts were added. This ensures that the Control Room switch must - be in the " start" or " auto" position in order for the engine to receive / maintain any start signal. Second, the generator output breaker auxiliary contacts were replaced with the output breaker's cell switch contacts. This eliminates the breaker "open" lockout in the start circuit. These changes do not affect the function of the HPCS DG with respect to its response to a plant emergency. The change also adds a relay in the start circuit which provides isolation of the lube oil recirculation pump, the jacket water keepvarm heater and the generator space heater. The prelubrication and prevarming functions of the isolated components is not required during the short period of time between receipt of any start signal and operation of the DG at rated-speed. Prelubrication is provided by the static pressure head from the lube oil cooler which is kept full during standby conditions by-the circulating pump. The volume of oil contained in the oil cooler is not affected significantly during this . period of time between pump isolatioin and engine operation hence prelubrication is not comprised. Categorically, the size of the heater for the generator and jacket water are small compared to the prevarmed mass of the engine and generator. Isolation of these heaters vould have no effect on the prevarmed condition of the DG over this short period of time. During engine operation, these heaters are not required to function for obvious reasons. Hence, isolation of these components during any engine operation vill'not impact the function of the HPCS DG. Therefore, the probability of the occurrence or the consequences of an accident or malfunctfon of equipment previously evaluated is not increased. II. No. All of the components installed as the result of this change meet all of the original system equipment qualification requirements. Additionally the physical installation of these components and their associated viring is in accordance with the original installation requirements. This change does not affect the function of the Division III DG. Hence, the possibility for an accident or malfunction different than previously evaluated is not created.
, Attachment 3 _
PY-CEI/NRR-1141 L Page 56 of 285 SE No.: 88-467'(Continued) III.~No. The margin of safety as' defined in the bases of Technical Specification Section 3/4-8 refers to the reliability of the on-site power supplies. As described in Items I and II above, the function of the Division 3 Diesel Generator has not'been altered by this design change. Hence, the margin of safety described in the Technical Specification bases is not affected.
7 l:j ,g Attcch:2nt 3
. j, - PY-CEI/NRR-ll41 L t
Page 57 of 285 SE No.: 88-468 Source Document: DCN 2462, Rev. 0
- Description of Change Add thermostat IN24-N0708 (installed per DCP 85-0099) to P&ID D-302-110 and its cor.nection to temperature element IN24-N200 on P&ID D-302-107. ,
These P&ID's correspond to USAR Figures 10.1-6, Sheets 1 and 4. s
. Summary b
I. No. This thermostat does not control any equipment related to the safe
. shutdown of the plant. The installation.vas evaluated under DCP 85-0099 and adding it to the drawings doesn't affect operation.
II. No. This thermostat does not control any equipment in the direct condensate stream. Failure to regulate cooling water to samples vill in no way increase the possibility of an accident. III. No. The samples that this thermostat regulates cooling _ vater to are not part of the Technical Specifications. p
> r
Attcchm:nt 3 PY-CEI/NRR-1141 L' Page 58 of 285 SE No.: 88-469 Source-Document: DCP 87-0524F, Rev.0 Description of Change This design change involves the installation of a 1/2 inch reactor water sample line for the Zine Addition (P85) System. This line branches off the Reactor. Water Cleanup (G33) system pump (1G33-C001A&B) vent and drain lines to a common point for analyzing reactor water zine levels. This change also involves the installation of a Nuclear Closed Cooling Vater (P43) System line to a P85 sample cooler.(1P85P001). This cooling vater line provides reactor water sample cooling prior to analysis. Summary
.I. No. This modification affects the nonsafety portions of the Reactor Vater Cleanup (G33) and Nuclear Closed Cooling (P43) Systems. The modification involves:
- a. The installation of a 1/2 inch reactor water sample line for the-Zine Addition (P85) System. The sample line branches off !
the existing vent and drain lines for each RVCU pump (1G33C001A and 1G33C001B) providing a common point for analyzing reactor water zine levels. In addition, the vent lines for the G33 pumps were relocated to a common high point on the P85 sample line, i
- b. The installation of a nuclear closed cooling vater line to a P85 sample cooler (IP85-P001). This cooling water line provides reactor vater sample cooling prior to analysis. .
Cooling water flow rate is approximately 3-4 GPM. The P85 modification to G33, does not' alter the function of the RVCU ; System as described in section 5.4.8 of the USAR. The consequences of a sample line break are far less severe.than the line break already evaluated for the RVCU System. The high energy portion of - the sample line is contained within the RVCU Pump Room. Thus, the high energy environmental effects of a sample line break in this room are enveloped by the G33 line break (Ref..USAR Section 15A.6.5.3d). That portion of the sample line passing into the RHR B Room (AB599-02) is considered lov energy (i.e., 150 psig, 105 degrees F.). The boundary exists at IP85-F025 located in the RVCU B Pump Room. In addition, this portion of the line is designed for low flow conditions (i.e., 400 cc/ min).
Attachn;nt 3 i PY-CEI/NRR-1141 L Page 59 of 285
.SE No.: 88-469 (Continued)
Summary The Nuclear Closed Cooling System is not required for safe shutdown of the reactor following an accident (Reference USAR Section 9.2.8). The P85 cooling line adds a negligible heat load to the Nuclear Closed Cooling System during normal operation (approximately 3-4 GPH) and, in the event of a line break results only in the loss of the P85 sample cooler. II. No. Since the P85 sample line provides only a process fluid route from G33 to an analyzer, a vorst case scenario is bounded by the G33 pipe break for Zone AB-5 and the E-12 pipe break for Zone AB4. Also, in accordance with Standard Review Plan 3.6 no postulation of the dynamic effect on high energy breaks in lines less than 1.0" dia is required. Thus, the P85 sample line is independent of any other system except those cited in Item #1. As a result, no possibility exists for an accident or malfunction of a different type than those previously evaluated in the USAR. Further, the postulated failur< modes due to a P85 system malfunction create no new threat to the reactor coolant pressure boundary or the fuel. At vorst, since the P85 system is essentially manually operated, a potential exists for over injection resulting in higher than desired levels of zine in the reactor coolant. This event is postulated to occur during the initial injection phase of zine addition in which reactor coolant zine levels would approach 10 ppb. The e' vent vould be associated with the loss of the P85 - reactor water zine analyzer during the mechanical injection of zine into the feedvater. However, the likelihood of this event occurring would be very remote and the effects so limited that no adverse impact would be incurred to the RCPB or the fuel. First, the unmonitored injection of zine although postulated would be extremely unlikely. Redundant grab sample analysis vould be provided via the P35 reactor water sampling system or the P85 feedvater analyzer. This redundancy precludes any uncertainty of zine levels in the reactor vater. Second, because of the extremely small amounts of zine injected (40 cc/ min), changes in reactor water zine levels would occur very gradually over a period of several hours. This vould provide more than ample response time to terminate injection in the event that an upward trend in zine concentration is detected. This mode of operation would further reduce the potential for over injection n to an improbable occurrence.
Attechtant 3 PY-CEI/NRR-1141.L Page 60 of 285 SE No.: 88-469'(Continued)
- Summary Third, a significant increase in reactor water Loluble zine (20 ppb) '
would be mirrored by a proportional rise in reactor water conductivity. Although soluble zine'in excess presents no adverse impact to BVR chemistry, measures would be taken under the existing-chemistry control program to maintain conductivity within fuel-
"Varranty Operational Limits". These measures vould include suspending zine injection.
Finally, excess levels of zine vould not disturb the oxygen-chloride relationship in the reactor water. As a result, an over injection of zine would not degrade the resistance of stainless steel components to intergranular stress corrosion cracking as described in section 5.2.3.2 of the USAR. Therefore, the presence of zine even in excess presents no threat to the reactor coolant boundary. - III. No. Since the likelihood of an occurrence or the consequences of an accident to either G33 or P43 resulting from the P85 reactor water sample line interface do not increase, the functions of these systems essentially remain unaffected. As a result, the margins of-safety as defined in the Technical Specification bases remain unchanged. t I
t Attachm:nt 3 - PY-CEI/NRR-1141 L
, Page 61 of 285 SE No.: 88-470 '
Source Document PAP 1910, Rev. 2 l
' Description of Change This evaluation examines'several changes:to PAP-1910, Fire Protection Program. Changes were made to the sections covering staff .
responsibilities, maintenance, fire brigade qualifications, training,: { and fire drills. Summary I. No. All changes made to this procedure vere evaluated with respect to the USAR. All changes made were found to be consistent with the - fire protection requirements of the USAR. Therefore, the probability _of an occurrence or the consequences of an accident previously evaluated in the USAR was not increased. t II. No.. All changes made to this procedure have been evaluated and found to be consistent with the fire protection requirements of the USAR. Therefore, the possibility of an accident different from any previously evaluated in the USAR has not been created. III. No. The Fire Protection Program is referenced under administrative' controls sections 6.5.1.6 N, 6.5.2.8 E, and 6.8.1 H of the Technical Specifications.- All changes made in this procedure vere evaluated and found to be consistent with the Technical Specification section- ' and the USAR. Therefore, no margin of safety has been reduced. l
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' Attach; Int 3 PY-CEI/NRR-1141 L' Page-62:of 285
' SE No.:- 88-471 Source Documents' PAP 1917, Rev. 2 Description of Change This evaluation examines several changes.to PAP-1917, Fire Protection Training Program. Changes'made to the sections covering. staff-responsibilities, fire brigade qualifications /requalification and training courses for various site personnel.
Summary ' See Safety Evaluation'88-470. SE No.: 88-472 Source Document: USAR CR 88-287 Description of Change 4 Evaluation-of'a.USAR change request to.Section 17.2.4.2 to delete a paragraph that makes reference to Professional Services Agreements, which' are no longer used to procure services for PNPP. Summary I..No. This is an administrative change and updates the USAR on procurement of services and consultants. It has no impact on the accident analysis .
, .II. No.- No accident or malfunction of a different type is created by this -
change. III. No. The margin'of safety for any Technical Specification is not -l affected.-
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'Page 63 of 285-SE No. . 88i473 .
Gource Documents- USAR CR 88-286 Description of Change Update USAR Figure 17.2-2 to reflect the most current NOAD Operational ! Quality Section organization change. Summary _I.-No. This is only an' organizational change / update of the USAR and therefore, cannot affect the probability of an accident or > malfunction of equipment.
.II. No. This change _is administrative and no accidents or malfunctions are involved.
III. No. Technical Specification Section 6.2.1 is not affected by this' change and no margin of safety is involved. L
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Attachm:nt 3
'PY-CEI/NRR-1141 L Page 64 of 285 SE No.: 88-475 Source Document: DCP 87-0305, Rev.0 Description of Change This design change installs two jack stations for communications in the Steam Tunnel. Each station is equipped to handle Public Address (PA-RSI), Maintenance and Calibration (R52) and Telephone (RSS). Also, an RS1 amplifier for the PA system is being changed to a handset / amplifier to enhance communications in the Turbine Building.
Summary I. No. The PA, Telephone, Maintenance and Calibration Systems provide communications for personnel in the Control Room and/or Emergency Shutdown Panel and work stations that may be required to be staffed following transients and/or accidents. Installation of two (2) PA and telephone jacks and maintenance & calibration jacks in the Steam Tunnel vill not increase the probability for a malfunction of the communication systems or other equipment important to safety as identified in the USAR. t II. No. Installation of PA, telephone, and maintenance & calibration jacks in the Steam Tunnel vill only affect the individual communication systems. Therefore, installation of this equipment vill not increase the probability for a different accident or malfunction to occur which has not been previously evaluated in the USAR. III. No.- Per review of Section 3/4.9.5 of the Technical Specifications no reduction in the margin of safety occurs as a result of this design change. SE No.: 88-476~ Source Document: PAP-1916, Rev. 3 Description of Change This evaluation analyzes changes made to PAP-1916, Duties of the Fire Vatch. Changes were made to the sections covering staff responsibility, fire watch responsibilities / assignments and fire hazards.
- Summary See Safety Evaluation 88-470.
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Attachrcnt 3 PY-CEI/NRR-ll41 L Page 65 of 285 SE No'.: 88-477 Source Document DCP 87-0276D, Rev.0 Description'of Change The design changes being made are to enhance the reliability of the . Division 2 Diesel Generator (R43) 125 volt DC control circuit. A common ' 125VDC control supply is-added which allows starting or stopping of the engine if either the "A" DC or "B" DC supply is available. An alarm vindov "125VDC Trouble" in panel lil51-P054B is being added to alert the operator of an abnormal condition of the DC power supply to diesel controls. An interlock is added (relays RA and RB) to the automatic starting of the diesel generator auxiliary systems M43, P45, R46, and R47 to prevent the systems from starting when the engine is in standby. Summary I. No. The changes being made by this DCP enhance the reliability of the , Division 2 Diesel Generator control and therefore the probability of a malfunction of this safety-related equipment vill be reduced. The control system enhancements are as follows: A common 125VDC control supply is being developed which will allow starting or stopping of the engine if either the "A" DC or "B" DC supply is available. This modification uses one less relay than the existing scheme and does not change diesel generator operation as described in the USAR. The elimination of a relay and the provision for start capability from either DC supply will enhance overall diesel generator reliability.
'The addition of an alarm vindov "125VDC Trouble" in panel lil51-P054B is being done to alert the operator of an abnormal condition of the DC power supply to diesel controls. This is being added to the diesel generator alarms as detailed in the USAR Page 8.3-33. This will enhance operability of the diesel generator and will not adversely effect diesel generator reliability.
An interlock has been added (relays RA and RB) to the automatic starting of the diesel generator auxiliary systems H43, P45, L R46, and R47 to prevent the systems from starting when the engine is in standby. This vill eliminate unnecessary stresses to the systems due to spurious supporting system initiations. These added relay contact interlocks are identical to other relay interlocks used throughout the diesel control circuitry c and their addition vill not adversely effect diesel generator l availability or reliability, i
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$,Y': III.,No; iThe margin'of saiety.-as' defined-in Technical Specifications Seetion.
', '3/4.8 Electrical Power Systems is not affected because. diesel ;
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Attzchrant 3 PY-CEI/NRR-1141 L
- Page 67 of 285 SE No.: 88-478 Source Document: DCN 2468, Rev.0 '
Description of Change Removal of associated instruments and root valves intended for use during the ASME Turbine Acceptance Test and replacement with pipe caps to prevent recurrence of previous piping failures, t L Summary I. No. No accident or malfunction of equipment important to safety is I postulated due to failure of small bore Main Steam (Nil) System piping downstream of the turbine stop and control valves. These , instruments were to be used during the ASME Turbine Acceptance Test
< and are not required during normal plant operation. The N11 system vill remain in compliance with USAR Chapter 10 following this modification.
II. No. There is no possibility for creation of a new kind of accident or malfunction as the new piping configuration conforms to ANSI /ASME B31.1 as required'in the USAR and as exists for the remainder of the system. III. No. Removal of instruments and root valves required for the ASME Turbine Acceptance Test does not affect the margin of safety in the Technical Specification Bases.' Calculated doses resulting from the maximum leakage allowance for the MSIVs in the postulated-LOCA r situations would be a small fraction of the 10 CPR 100 Guidelines, provided the main steam line system from the isolation valves up to and-including the Turbine-Condenser remains intact. Removal of the subject valves and instruments vill improve the integrity of. the Main Steam System. l l l b
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, r Attechnent 3 PY-CEI/NRR-1141 L 4
Page 68 of 285. SE No.: 88-479
' Source Document: DCN 2448, Rev. O Description of Change The function of the startup strainers was to ensure protection of downstream equipment during initial startup. It is a good practice to leave these strainers in place during the operations phase as added protection. P&ID 302-081 was revised to denote these as permanent strainers, via this DCN. All the strainers are located within the Heater.
Bay Building. The strainers are upstream of the following components: Strainer Upstream Component 1N27-D004A & B Hain Turbine Feed Pump A and B IN27-D005 Hotor Driven Feed Pump 1N27-D006A & B Recirculation Minimum Flov Valve F160A and B 1N27-D007 Recirculation Minimum Flov Valve F170 { Summary
.I.-No. USAR section 10.4.7.2.4 (safety evaluation) addresses the Feedvater ,
System. This section does not address strainer failures within the J Feedvater System. The section does address specific component ' failures, and they are as follows:
- a. One reactor feed pump failure-1
- b. Feedvater heater tube rupture fi
- c. Guillotine break of a feedvater header out side of Containment, !
l reference Chapter 15 of USAR. The only components in the Feedvater System which are important to safety are the feedvater outboard / inboard containment isolation i valves. This is addressed in the USAR under the specific component failure mentioned under letter c. above. The strainers identified j do not affect / impact these safety-related valves nor are the strainers near these valves. These strainers have been in operation since preoperational testing and have not failed. Therefore, the probability of occurrence or the consequences of an accident or ! malfunction of equipment important to safety previously evaluated in ; the USAR has not increased.
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Attachz:nt 3 k PY-CEI/NRR-1141 L-Page 69 of-285 SE No.: 88-479 (Continued). Summary *
= II. No. As mentioned previously these strainers vould protect the equipment downstream. The Feedvater Piping was sized with these strainers in mind and has minimal pressure drop across them. Vith the strainers
-in line the reliability of the equipment downstream would'be increased. The strainers have been in operation since preoperational testing and have not failed. If a strainer were to fail it vould fall at the screening basket.
This type of failure vould be highly unlikely to occur. If it were to occur, the screening basket would follow the piping and get lodged in an elbow or at the existing reducer upstream of the pumps. It would be highly unlikely that the strainer would pass the elbov/ reducer if a failure of the strainer would occur. If it were to occur-the strainer vould enter the feed pump suction and impact on the pump's impeller. This vould cause damage to the pump and would require it to be isolated. The feed pump failure has been addressed in the USAR.- On the other hand, the strainers which would L enter the minimum flow recirculation valve inlet vould lodge in the inlet.of the, valve because of the valves internal configuration. This vould cause the valve to be inoperative. These valves are used for'startup purposes and would not impact plant safety. The possibility for an accident or malfunction of a different type-than ' any evaluated previously in the USAR has been reviewed and determined not to exist. III. No. The strainers identified are not addressed in the Technical Specifications. l 1 l 1 l
F L Attach ;nt 3~ PY-CEI/NRR-1141 L i Page 70 of 285 i SE No.: 88-480 > Source Document LL&JED 1-88-186 Description of Change This LL&JED temporarily modifies the Reactor Recirculation Pump B Carbon Dioxide Fire Suppression Control-Panel (1H51-P781) from dual-zone actuation-to single-zone. actuation. ' Summary I. No. This LL&JED provides for temporary modifications of the Reactor
, Recirculation Pump "B" Carbon Dioxide Fire Suppression Control Panel lH51-P0781 from dual-zone (cross-zoned) heat detector activation to-a single-zone activation. The modification vill not increase the probability of occurrence of a fire in the Reactor Recirculation Pumps, and vill not increase the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR in that the installation of this LL&JED will provide for a means of maintaining the capability of initiating carbon dioxide fire suppression in the event of a fire in the Reactor Recirculation Pumps.
s The LL&JED is being installed due to an apparent ground on one of the two heat detection strings for Recirculation Pump "B". The , ground is located inside of the dryvell, and is not accessible for ' troubleshooting during plant operation. By lifting the detection string terminations in the control panel, the ground is removed from the control panel allowing for reliable operation of the remaining detection string, associated functions, and carbon dioxide activation. In addition, the control panel for Recirc Pump "A" (lH51-P212) is located within a common panel enclosure utilizing a common power supply. The climination of the ground increases the reliability of 1H51-P212 to perform its design functions. Inside of the dryvell, (USAR Fire Zone IRB-le), redundant safe shutdown circuits and equipment are either spatially separated or provided with radiant energy heat shields to prevent a fire from damaging both divisions. l II. No. The fire hazards analysis (USAR Appendix 9A) describes the effects of a fire in the drywell (Fire Zone IRB-1c). This HPI/LL&JED vill-not create.the possibility of an accident or malfunction of a different type than previously evaluated in the USAR. III. No. Only administrative and audit aspects of fire protection are in y Technical Specifications. : 1
Attach:;nt 3 PY-CEI/NRR-1141 L Page 71 of 285 .SE No.: 88-481 Source Document: USAR CR 88-292 Description of Change Evaluation'of a USAR change request to Sections 11.2.1.8 and 11.2.2.12 to
. eliminate statements regarding the use of an automatic mode on the radvaste Programmable Logic Controller (PLC). Testing of this function is being eliminated because operation in this mode is not going to be pursued because of the numerous cross connects between vaste streams.
~ Summary i I. No. This change request was generated as a result of an Activity Value Analysis (AVA) suggestion to eliminate the automatic function of the Liquid Radvaste (LRV) Programmable Logic Controller (PLC). The only protective feature lost in eliminating this function is the automatic prevention of tank overflow. However, because all tanks in the LRV system have closed tops with overflow lines piped up to solid embedded drain piping that is routed to sumps and then back to the LRV collection tanks, the consequences of tank overflow remain unchanged. Because tank overflow is 100% contained by the system described above and the LRV tanks are either housed inside reinforced concrete structures with floor drains for routing any leakage to the LRV system or within seismically qualified retaining structures,the probability of occurrence of an accident previously analyzed in the safety analysis report has not been increased. The design provisions to prevent uncontrolled release of radioactive material to the environment as a result of any single malfunction or operator error is not impacted by the elimination of this automatic function. USAR Table 11.2-1. details the evaluation for each specified component. failure. This change does not negate anything in this Table, nor does it modify any assumptions made in USAR Chapter 15. II, No. The event that could occur with elimination of the PLC is tank overflow. The LRV system has been analyzed with respect to tank rupture, equipment malfunction and small leaks in the system process lines that transport liquid radvaste. Therefore, the possibility for an accident or malfunction of a different type than any previously evaluated has not been created. III.1 No. Technical Specification 3.3.7.9 and associated tables take no credit , for the PLC. This controller does not impact any other Technical Specification or margin of safety assumed in the Technical Specifications.
y y, lif ; ' .' n ' Attrehm:nt 3 PY-CEI/NRR-1141 L Page 72 of 285 SE No.: 88-482-Source Document: 0H16P Physical Security Plan, Rev. 12 Description of Change i Revision 12 of OM16F Physical Security Plan, Rev. 12 has been evaluated ! to ensure that the effectiveness of the Perry Nuclear Power Plant ' Security Plan has not been reduced and to ensure that these changes meet the requirements of the 10 CFR, Part 73, Physical Protection of Plants f: and Materials. Site Protection must be contacted for further details because information is considered " safeguarded." Summary
'4 I. No. OM16F describes the comprehensive Physical Security Program ano therefore, does not effect the' occurrence or consequences of an
- g. accident or malfunction of equipment. j II. No. OM16F does not direct the operation of plant systems or equipment and, therefore, does not created the possibility for an accident or malfunction..
III..No. OH16F does not reduce the margin of safety as defined in the basis for any Technical Specifications, i l l l i l 1 f .,
e j . 1 Attachm nt 3 PY-CEI/NRR-ll41 L ! Page-73 of 285 SE No.: 88-483 Source Document: DCN 2478 , Description of Change This change revises delta-P (op) instrumenta' ion setpoints and airflow values relating to the Annulus Exhaust Gas Treatment (AEGTS-MIS) System to maintain consistency within the design bases. Summary I. No. The original design bases and safety analysis requires the secondary containment (annulus area) to be maintained at a minimum negative pressure of 0.25 inches water gauge at all times. The design approach used to recalculate the setpoint for the op instrumentation remains consistent with the original design bases and safety analysis and accounts for the phenomenon described in NRC - Information Notice 88-76 with adjustments for specific conditions at , the Perry Plant._ The op and airflow values being revised vill permit the AEGTS to operate and maintain secondary containment integrity as described in the USAR. Based on the fact that the overall system function has not changed and the parameters upon which the analysis (Section 15.6.5.5.1.2a) in the USAR was based have not been affected. Therefore, the probability for the occurrence or the consequences of an accident or malfunction of equipment previously evaluated is not increased. II. No. Revision of the op setpoint will permit the system to function as described in the USAR. The change does not affect or change overall system function. Therefore, malfunctions of a different nature vill not be created.
.III. No. The Technical Specification addresses secondary containment integrity. The setpoint change being requested remains consistent with the bases. The change of the negative pressure maintained has moved in the conservative direction and is administratively controlled to meet the technical specification intent. System function remains within the parameters originally specified and has not changed. Therefore, the margin of safety as specified in the Technical Specifications has not been reduced,
y --- _ AttcchnInt 3 PY-CEI/NRR-1141 LL Page 74 of 285 SE No.: 88-485 Source Document: SXI-0033 Description of Change This_special-test instruction describes a method for locating leaking, fuel bundles using flux tilt. All required rod motion is performed using applicable instructions and policies described elsewhere in the
-Operations Manual, Summary I. No.' Rod motion vill be within pre-analyzed limitations. No limits vill' be bypassed or altered. Thus, the probability and consequences of accidents will be unchanged.
II. No. All rod motions will be in accordance vith the required design and administrative limitations, and other equipment used is merely monitoring.~ III. No. All rod motion vill be in accordance with Technical Specification requirements.- i 1
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Attachnnt 3' PY-CEI/NRR-1141 L g Page 75 of 285 g SE No.: _.88-486 Source Document: DCP 87-0524G, Rev.0 Description of Change This design change provides for the installation of a feedvater sample line to the P85 feedvater zine analyzer. This sample line branches off i an existing sample line for the Turbine Plant Sampling System P33. 'j
. Summary
- 1. No. The modification described in this design change ~affects the P33 system. l 1 ,
The modification involves tapping off the final. 'l feedvater sampling line (AP-37, D-302-181) to provide a sampling route for the Zine Addition (P85) System. The addition of the P85 sample line modification _to P33 does not i alter the function of the Turbine Plant Sampling System as described 1 in Section 9.3.2 of the USAR. The addition of zine to both the reactor feedvater and reactor coolant _vas previously evaluated and determined to create no i accident or malfunction possibilities (Reference Safety Evaluation 88-469). The vorst case scenario is limited to a 1/4 inch sample line break. Under Standard Reviev Plan 3.6 no dynamic effect need be postulated for high energy lines less than 1.0 inch in diameter. , l This makes the P85 feedvater sample line independent of any system. ' Also, because of the extremely lov flow-(< 400cc/ min) and the sample. ; line branch being downstrcam of an existing P33 sample cooler, a l line break would result only in a manageable spill. Therefore, the ' probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in ' the USAR is not increased. i II. No. See Item I above. III. No. Since the P85 sample line presents no adverse effects on the operation of the P33 Turbine Plant Sampling panel, any margins of safety as defined in the Technical Specifications remain unchanged. l l l
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Attccht:nt 3 PY-CEI/NRR-ll41 L i
, Page 76 of 285~
SE No.: 88-487 - Source Document: DCP 87-0654, Rev.1
- Description of Change This design change removed the turbine low pressure stop valve limit-switch from the Feedvater Control (C34) Syetem for both Reactor Feed Pumps.
1 Summary I. No. The original design intent of the turbine trip logic in the C34 Feedvater Control System was to provide pump status to the C34 gain change circuitry. The present configuration of_the RFP turbine trip logic utilizes three sets of contacts. The first contact opens when RFP discharge valve is less than 90% open. The second contact opens when loss of hydraulic pressure to the RFP occurs. The third contact opens when RFP low pressure valve is not fully open. t The RFP low pressure stop valve requires periodic testing as recommended by GE. This is a standard practice throughout the industry. While cycling the valve on 9/9/87, the low pressure stop r valve contact opened and gave a " false" ' indication of an RFP trip i which eventually resulted in a Reactor Scram. Therefore, a design deficiency was noted and the issuance of EDCR 87-0654 was made. After review and analysis, the original work scope of EDCR 87-0654 ! included the use of the RFP control valve limit switch instead of the low pressure stop valve contact. The use of the control valve contact would maintain the present design configuration. DCP 87-0654 was issued and sent to the plant for review. After further review of DCP 67 -0654, it was determined that two possible' "falso trip" conditions e?ist when using the RFP control valve limit switch. These " false trip' conditions'can occur when the C34 , control system automatically responds to a decreasing flow demand or the setpoint setdown circuit engages. The control valve limit i switch is commonly used as turbine trip status at other plants. The Perry Plant has a unique situation which can cause the above " false trip" conditions. The Perry RFP has a rapid (faster than the average) responding control valve which can momentarily close when not necessarily in a trip situation as stated above. Thus it is not
=
prudent to use the RFP control valve limit switch, i 1 l l l
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Attach:;nt 3 PY-CEI/NRR-1141 L 4 Page 77 of 285
.SE No. 487, Rev. 1 (Continued)
Summary - Engineering has chosen (with GE concurrence) to have no RFP (control or stop valve)-limit switches in the C34 turbine trip logic. This is because existing hydraulic trip system low pressure switch and ' RFP discharge valve limit switch provides more than adequate RFP pump status.to the C34 System. :r Therefore, DCP 87-0654,-Rev. 1, vill still meet the original design intent.and-improve system reliability by eliminating " false RFP ' trip" signals to the C34 Control System. . The C34 Control System-vill operate as originally designed and does not affect any safety systems, t II. No. No new failure modes have been created since the original design- < intent is~still met and operation has been improved by eliminating -
" false trip" indications.
III. No. This item does not affect the Technical Specifications thus, the margin of safety has not been reduced. l l l 1
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-Attcchmtnt 3 1 PY-CEI/NRR-1141 L 1 Page 78 of 285 i SE No.: 88-488-88-488, Rev. 1 Source Document: DCP 87-0628B, Rev. 0 DCP 87-0628B, Rev. 1 1
-Description of Change Relocate and enlarge to 2 inches the Instrument Air (PS2) System supply lines to valves IN27-F0160A and B, IN27-F0170 and 1N27-F0307 to allow for
-the additional air flow required by the new Feedvater (N27) System " Drag" valves. Fabricate new lines from stainless steel.
Revision 1 of this safety evaluation reviews the effect of the change of scope of'this DCP since 3 of the 4 valves and corresponding flex hoses, are not being installed during the first refueling outage. Summary I. No. .The ncv N27 valves require an increased air supply over what was previously available through the existing J-manifold. The addition of the PS2 instrument air piping in the heater bay to support ; N27 valve operations increases power plant safety by providing an-adequate air supply to the N27 valves. The function of'the air supply to the N27 valves remains the same as in the existing design. Therefore, the probability of occurrence or the consequences of an . accident or malfunction of equipment important to safety previously- ' evaluated in the USAR is not increased. In addition, overall reliability of the N27 valves is increased, from an air supply I quality viewpoint, as.the DCP installs local air filters to ensure reliable valve operation. II. No. The function of this portion of the Instrument Air System remains-the same as.the existing piping, there can be no new 3 accident / malfunction possibilities created by this DCP. Ill. No. The margin of safety designed into the components / equipment identified in the Technical Specifications is not decreased as this DCP increases the reliability of the new N27 valves by ensuring that an adequate supply of good quality instrument air is available to the new N27 valves. This DCP does not change any information in the - Technical Specifications or alter previously described requirements.
Attachm:nt 3 PY-CEI/NRR-1141 L Page 79 of 285 t SE No.: 88-489 ' Source Document: DCP 87-0667F, Rev. 0 ' Description of Change
- This design change adds an addition to the Service Building to house a Vertical Boring Machine. 'This DCP also-provides for the machine's installation. ,
Summary I. No. The addition to the existing Service Building is nonsafety-related and vill house the nonsafety-related Vertical Boring Machine. The
-11' x 31.5' x~18.5' high (approximate size) addition is located east of the existing stair tower on the south side of the Service Building, and vill be shown on the revised Figure 1.2-5 of the USAR.
The addition is structurally adequate for design loads per calculations in file code 11:29 Rev 0. The additional loado on the existing Service Building vere evaluated and found acceptable (Reference File Code 11:04 Rev. 3). Therefore, the probability of an accident / malfunction of equipment important to safety is not increased. II. No. The addition was designed for the applicable design load conditions
-(including tornado vind) and combinations, therefore the possibility of an accident / malfunction of a different type is not created.
III. No. The structural addition to the existing Service Building does not affect the Technical' Specification margin of safety since this is not a functional-change.
e I Attachrent 3 PY-CEI/NRR-1141 L Page 80 of 285 SE No~.: 88-490
-Source Documents- FTI B09, Rev. 1
. Description of Change
'This evaluation analyzes the method of recovering from out-of-position control rods as described within the Reactor Engineering Instruction FTI,B09, Recovery of Out-of-Position Control Rods.
Summary. I. No. This instruction provides directions for bypassing the Rod Control & Information System (RC&IS) inhibits and moving control rods.for the purpose of recovering from control rod maloperation events without increasing the risk of a Control Rod Drop Accident (CRDA) or the Rod Vithdrawal Error (RVE) in a manner that does not exceed the MCPR_ limits and that is-consistent with Technical Specification during events which have the potential for exceeding the MCPR limits.
.II. No. Since this instruction only provides directions for bypassing the RPCS inhibits in order to recover from a control rod maloperation, the only events which need to be evaluated are those for which the RPCS is designed, the Control Rod Drop Accident and the Rod
.Vithdrawal Error, j
III. No. Since this instruction provides directions for bypassing.the RPCS inhibits in order to recover from control rod maloperation in a manner that does not exceed the MCPR operating limit and that is .. 1 consistent with the Technical Specifications during events which , have the potential.for exceeding the MCPR operating limit, the margin between the MCPR operating limit and the MCPR safety limit is j not reduced. I
.. =
4
p m Attacht:nt 3 PY-CEI/NRR-1141 L Page 81 of 285 < SE No.: 88-491-Source Document: DCP 86-0926, Rev. O t Description of~ Change The existing Main Steam Isolation Valve (MSIV) Leakage Control (LCS-E32)
, System flow measurement system has experienced numerous problems. The system vill be replaced with probe type, IE qualified flov transmitters-from'FCI. (I&C/ Electrical Evaluation) r Summary I. No. The HSIV Leakage Control System is used to control the release of' i radiation after a LOCA has occurred. The existing flow measurement !
portion of the system consists of a LVDT type flow element and transmitter. This sub-system monitors the steam flow through the '
'HSIV LCS piping. If the flov is not lower than the setpoint in a ;
certain time frame, then the leakage control system (upstream) vill isolate. The change in flow measurement instrumentation vill not change'the system function. The existing flow transmitter and flow elements will be removed. The new qualified probe type flow transmitters vill be installed in the same location as the original transmitters. An orifice vill be installed to provide proper flow restriction. 3 The new transmitters have a greater reliability since they are not ~ damaged by water in the system. They are more accurate and easier-to calibrate than the existing-type because they have a linear output. Thus, the new flow transmitters will not' increase the probability of a malfunction,, but vill make the HSIV Leakage-Control System more reliable. II. No. Since the flow measurement system function has not changed, the possibility for an accident different than previously evaluated is not created. III. No. The flow measurement system must be operable per Technical Specification 3/4.6.1.4. The replacement of the existing flow transmitters with FCI flow transmitters, vill make the system even more reliable. Therefore, replacement of these transmitters will not cause a reduction in the margin of safety.
x _. y IJ Attcch::nt 3 L PY-CEI/NRR-1141 L [' Page 82 of 285 SE No.: . 88-492-
, Source Documents. .DCP 86-0926, Rev. 0 Description of Change Design change is same as described in Safety Evaluation 88-491. However, a mechanical evaluation vas'also performed since the old flow elements vill-remain in place to act.as flow restrictors. (Mechanical Evaluation)
Summary I. No. Nev' instrument with separate orifice is functionally equal to existing design with orifice internal to instrument. II. No. .See Item I above. III.'No. See Item I above. i w _____m., _ _ _ _ - _ _ _ _ _ _ _ _ _ . - _ - - - _ - - . - - . - - - - - - - -
Attachment 3 PY-CP1/NRR-ll41 L Page 83 of 285 SE No.: 88-493 Soutoc Document: USAR CR 88-284 i j Desetiptson of Change Evaluation of a USAR change request to Table 6.2-40 which adds Note 13 to penetrations P131, P204, P310, P311. P404 and P405; indicating that the corresponding systems are not vented and drained for Type A testing. Additionally, valves E22-P0517 and F0518 are deleted from Table 6.2-40.
- Summary I. No. 10 CFR 50 Appendix J. 211.A. Type A test states the containment atmosphere is allowed to stabilire for a period of about four hours after teaching test pressure prior to the statt of the Type A test.
The containment cooling systett- and the chilled vater system are sun, as necessary p f or to and during the Type A test to maintain stabilized containment atmosphetic conditions This would require penetrations P310, P311, P404 and P405 not to be vented. The Contiol Rod Drive (C11) System should have a small amount of flow to prevent debris from settling in the guide tubes. This vould requite the C11 penetration P204 not to be vented. The draining r.nd venting of G33 penetration P131 vould leave one isolation valve between the teactor vessel and drain. This valve vould be in accessible du ing the ILRT. The eiore, G33 penetration 131 should not be drained. The removal of the two test connections vill not change the probability of an evaluated accident since the valves are verified locked closed and capped by SVI-T23-T1201. This instruction was used as the ienson fot eliminating the ether test connec tions !:c:. the table. Further, Appendix J states that these penettations vill be tested using a Type C test.
- 11. No. See Item I above.
111. No. Not venting these penetrations during performance of a Type A test does not change the margin of safety as defined in Technical Specifications since a Type C test vill he run on the penetrations in question. The deletion of the test connections has already been made to the Technical Specifications.
Atttchment 3 PY-CEI/NRR-1141 L Page 84 of 285 SE No.a. 88-494 { Source Document: DCP 88-0079, Rev. O l i Description of Change ! This design change implements changes to correct various Human Engineering Deficiencies (HED). HED-33 originally required yellow l i annunciator berels to be installed on certain Control Room annunciator i vindows. Operational experience reviews indicate that the annunciator ! demarcation is too distinctive and of no real benefit to the operators, ; and in some cases has caused confusion. It is felt that the regrouping of the annunciators into functional groups combined with the use of
-i priority color coding is acceptable and that the annunciator demarcation should be deleted' *
[ Summary !
- . No. Removal of yellow annunciator berels is a cosmetic change and does not impoet the safety analysis described in the USAR.
{ II. No. Removal of yellow annunciator bezels affects is a cosmetic change ( and cannot erente a new accident or malfunction. l 111. No. Annunciator vindow color is not described in the Technical i Specification bases. I f
-I I
i n t b l l w ,+ -. , w . - - - . w - w --%-.
Attcchment 3 PY-CEI/NRR-ll41 L Page 85 of 285:
-TE No.: 88-495 Source Document: SCR l-88-0085 through, 1-88-0088 I
Description of' Change Setpoint changes to various Unit 1 Area Radiation Monitors to eliminate Control Room nuisance alarms. High and alert alarm setpoints will be revised to reflect increased normal background radiation levels. Twice 3 background for alert and three times background for the high alarm as described in the USAR Chapter 13 Section 12.3.4.1.2.g. [ Summary 1 I. No. Radiation monitor setpoints provide an informational function only. They have no impact on the increased occurrence of any accident.
'II. No. Radiation monitor setpoints provide an informational function only.
.They have no impact on the increased occurrence of any accident.
III. No. Radiation monitors described are not contained in the Technical Specifications, therefore, the margin of safety as described in the basis of Technical Specifications is not reduced.
1 Attach::nt 3 PY-CE1/NRR-ll41 L Page 86 of 285 SE No.: 88-496 Source Document: DCP 85-02951, Rev. O Description of Change This design change installs power and contrcl circuitry for the charcoal filter unit used in the small tool decontamination area. Summary I. No. This design change installs pover, and control instrumentation circuitry associated with the small tool decontamination / storage area's charcoal filter unit. A portion of these circuits are required to support operation of the fire protection / detection system of this filter unit. This DCP does not install all of these circuits in accordance with NFPA 72D-Int: 311ation, Maintenance and use of Proprietary Protective Signaling Bvstems in thatt
- a. The Hi-Temp alarm, ill-lli-Temp alarm and drain valve actuation solenoid are not supervised.
- b. A fire protection system troubic alarm for panel H33-N141 is not annunciated in the central supervising station,
- c. The power supply for panel H33-N141 is not povered in accordance viti NFPA-72D.
The above deviations to NFPA 72D do not adversely affect the approved fire protection program in thatt
- a. Periodic Tests / Inspections vill be established to verify the operability of the fire detection / protection circuitry associated with this change. These inspections are adequate to ensure system operability considering that a high temperature condition is annunciated in the Control Room by three mechanisms (hi alarm, hi-hi alarm and analog temperature signal). The loss of any one of these signaling lines vill not prevent the operator from receiving sufficient information to recognize that a condition exists which may require the actuation of the fire suppression system. Failure of more than one of these circuits in the period between inspections is not credible. Supervision of the drain solenoid circuit is not required due to its short run.
- s k L Attachment 3 PY-CEI/NRR-1141 L I Page 87 of 285 SE No.: 88-496 (Continued) j i Summary I
- b. Loss of power to panel M33-N141 is not directly annunciated.
This is acceptable since loss of power to the load center which ! feeds it (load center FID) is annunciated in the Control Room utilizing circuits backed up by the non-divisional station batteries and the diesel generators. In addition, the ! intentional removal of power from any of the power supplies which feed M33-N141 which are down stream of FID will be i administrative 1y controlled. Considering that unintentional removal of power to M33-N141 is not likely and that intentional removal of power from M33-N141 or any power supply feeding it vill be administratively controlled, direct loss of power indication for panel M33-N141 , is not required. Compensatory measures vill be established when loss of power to M33-N141 is detected or removed intentionally. Therefore, the probability of occurrence of an accident or malfunction of equipment important to safety. previously evaluated in the USAR is not increased. The consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not ' increased. II. No. This change does not adversely affeet the Fire Protection Program; therefore, the possibility of an accident or malfunction or a different type than evaluated in the USAR is not created. III. No. The only aspect of the Fire Protection Program addressed in the basec section of the technical specifications is alternate shutdown capability. This change does not adversely affect alternate shutdown capability therefore, the margin for safety defined in the bases section of the Technical Specifications is not decreased.
Attcch::nt 3 PY-CEI/NRR-ll41 L Page 88 of 285 SE No.: 88-497, 88-498, 88-502 Source Document: DCP 87-0276A, Rev. O Description of Change This evaluation analyzes design changes made to the Division I Diesel Generator (DG). The change include modifications to the DG Control Circuitry, replaces the lube oil and jacket keep varm heater temperature control switches, and adds an interlock to the control circuits of various DG auxiliary systems to prevent unplanned cycling when the engine is in standby. Summary
- 1. No. The changes made to the DG Control Circulty include modifying the engine speed reset and adding the capability to interrupt engine
" cranking".
The portion of this design change which affects the engine speed reset eliminates the pneumatic control system interface upon receipt of a bus undervoltage or LOCA sigtsal, and provides for speed reset in the event the engine is already operating and either signal is received. This is a functional improvement from the original design which would only reset speed upon receipt of a LOCA signal. This ensures that the Diesel Generator vill perform its intended design function. This change increases overall reliability of the DG due to the elimination of dependence on control air for engine speed reset. This design change provides the capability to interrupt cranking of the engine with the addition of "inop" relay contacts in the control system. Thus in the event of a malfunction during cranking, the operator vill have the ability to interrupt cranking and correct the malfunction before all starting air is consumed. The "5-start" capability of the DG is not compromised by this change. This design change replaces the existing nonsafety-related temperature switches which control the lube oil and jacket vater heaters with a different design of temperature switch. These temperature switches do not directly impact the DG control system. The switch installation is consistent with original installation standards. In the event of a power loss to either of the DG tachometer transmitters, the relay contacts in the affected transmitter change state. As a result, this loss of power causes the unplanned initiation of engineered safety features, such as the associated DG Building Ventilation and Emergency Service Vater System components receive automatic start signals. This design change eliminates these inadvertent actuations with the implementation of a DG start
1 I Atttchment 3 i PY-CEI/NRR-1141 L : Page 89 of 285 : SE No.: 88-497, 88-498, 88-502 (continued) ! i Summary ^ signal interlock. .This interlock has no other impact to the evaluated design of the DG control system with respect to normal or
- emergency operation. Therefore,'the probability of occurrence or ,
the consequences of an accident or malfunction of equipment j important to safety previously evaluate is not increased. l r II. No. .This modification does not alter the operation of the Diesel Generator. . Further, all of the new components used in this : modification meet all of the original equipment qualification requirements. Hence, the possibility of an accident or malfunction of equipment of a different type than evaluated previously is not created. . i III. No. Technical Specifications Section 3/4-8 refers to the reliability of . the onsite power supplies. As described in Item I and II above, the operability of the Division I Diesel Generator has not been changed. ' Hence, the margin of safety described in Technical Specifications has not been reduced. ) I i i f h b P t b I
Attachment 3 PY-CE1/NRR-1141 L Page 90 of 285 SE No.: 88-499 Source Document: DCP 87-0633, Rev. O Description of Change This design change adds a 6-inch pipe bypass around Residual Heat Removal (E12) System valve 1E12-F024A for Suppression Pool Cooling / Cleanup (SPCC). It also adds new valves 1E12-F609, 1E12-F610. Operating License Amendment 21 modified plant Technical Specifications before the implementation of this DCP. (Mechanical Evaluation) i Summary I. No. The new piping and valves is considered moderate enargy piping, Class 2, break exclusion up to valve 1E12-F610. ASME III, Class 2 and 3 vill be utilized in the modification / installation of piping and valves. Pressure boundary integrity is assured with a minimum of 300 psi rated valves. Two new outside containment automatic isolation valves in series with an existing stop check valve. The two new valves vill have separate power supply via Division I and Division II service. The combination of the two automatic outside containment isolation valves, along with the piping inside containment submerged below the Suppression Pool, is an acceptable alternative to General Design Criterion 56, per ANS 56.2/N271-1976. Type C leak rate testing of containment valves (per 10CFR50, Appendix J) ensures that the total containment leakage volume vill not exceed the safety analysis assumptions at peak accident pressure. In addition, these new valves (1E12-F609 and IE12-F610) are to be classified as "normally closed" and opened only during operation of the SPCC return path to the Suppression Pool. The consequence of one of the containment valves failing to close on RPV Level 2 does not prevent Lov Pressure Coolant Injection (LPCI) "A" loop, Containment Spray "A" loop, or Suppression Pool Cooling "A" loop from performing their intended design function. Containment integrity is maintained with the two isolation valves, if one was to fall open. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. See Item I above. III. No. The proposed change vill preserve the environment in the event of a release of radioactive material to the containment atmosphere. The standard closure criterion for the new valves and appropriate leak rate testing preserves the assumptions of the limiting LOCA evaluation to ensure that offsite dose consequences are not increased. Therefore, this change does not produce a reduction in safety saargin contained in any Technical Specification, l 1 1
r i I Attcchment 3 PY-CEI/NRR-1141 L Page 91 of 285 i SE No.: 88-500 Source Document: DCP 87-0633, Rev. 0 , Description of Change This design change adds a 6-inch pipe bypass around Residual Heat Removal (E12) System valve IE12-F024A for Suppression Pool cooling / Cleanup. It also adds valves IE12-F609, IE12-F610. (Fire Protection Evaluation) Summary I. No. This change adds valves which affect the Safe Shutdown Capability Report (SSCR) which is part of the approved fire protection program as referenced in the USAR. These changes are implemented in accordance with Appendix R requirements, therefore
- a. The probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased,
- b. The consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No. This change does not adversely affect the fire protection program; therefore, the possibility of an accident or malfunction of a different type than evaluated in the USAR is not created. III. No. The only aspect of the Fire Protection Program addressed in the bases section of the Technical Specifications is remote shutdown capability. Remote shutdown capability is not required for a fire in fire area IAB-lb, which is the only plant area affected by the mechanical aspects of the addition of valves E12-F609 and E12-F610. t _ . , I
Attachment 3 i' PY-CEI/NRR-1141 L i Page 92 of 285 i l SE No.: 88-501 Source Document: DCP 88-0167, Rev. 0 Description of Change This design change involves the re-tubing of the Emergency Response ) Information (C95) System input signal-from the alternate level controller l to the manual level transmitters for the high pressure heater drains and 't vents.
-Summary
'I. No. There is no increase in the probability of occurrence or consequences of an accident or malfunction of equipment important to-the safety previously evaluated since the same design and g
installation parameters as the original design vere taken into account.in the evaluation of this installation. Further the system analysis has shown that failure of the high pressure heater drains and vents vill not compromise any safety-related systems or prevent safe shutdown. II. No. See Item I above. III. No. This installation vill not affect the margin of safety defined in the bases of the Technical Specifications since the system functionability has not been changed or reduced. This system serves no safety function. i
.)
L__.
~)
Attechment 3 PY-CEI/NRR-1141 L Page 93 of 285 SE No.: 88-503 Source Document: SCR's 1-88-1585 through 1-88-1588 Description of Change These setpoint change requests decrease the liigh Containment Pressure-RHR Spray Initiation setpoint and allovable value in the conservative direction and add analytical limits in units of psia for completeness. It also adds leave-as-is-zone and reset calibration information to the setpoint list. Summary
- 1. No. USAR Section 6.2.1.1.5.4 indicates that containment spray is initiated after containment pressure reaches 9.0 psig. USAR Section 7.3.1.1.4 describes the conditions which must exist for containment spray to automatically initiate, which inclurles the criteria that containment pressure must equal or exceed 9 psig. Both USAR sections are consistent in design description based on 9.0 psig being the analytical limit (process parameter) by which RilR spray should actuate. The proposed changes to the trip instrumentation setpoint and allovable value are being made in the conservative direction with respect to the 9.0 psig Analytical Limit. These parameters are derived from the analytical limit to account for instrument accuracy, calibration accuracy, and drift, thus ensuring trip actuation prior to or at the analytical limit value. Based on the desire to expand the leave-as-is-zone by a specific amount, it is required that the Setpoint and Allovable Value be adjusted downward.
The addition of the analytical limit to the setpoint list in units of psia vill be consistent / equivalent to the USAR specified 9.0 psig value. Addition of leave-as-is-zone and reset parameters to the setpoint list is for calibration purposes only and does not affect the required trip function described previously. Based on the discussion provided above, it is concluded that the probability of occurrence or the consequences of an accident / malfunction of safety-related equipment previously evaluated in the USAR is not increased. II. No. The scope of this evaluation is limited to the setpoint parameter changes / additions described above. A different type of accident or malfunction than evaluated previously in the USAR is not considered to be created. III. No. The changes described above are conservative with respect to the high containment pressure setpoint and allovabic value specified in Technical Specification Table 3.3.9-2. The margin of safety as defined in the bases for Technical Specifications is therefore not reduced. 1 1
; :r Attechment 3 PY-CEI/NRR-ll41 L !
Page 94 of 285 SE No.: 88-504 i sc Source Document: USAR CR 88-271 i Description of Change ; This change adds a statement to clarify that certain Section 3.9 tables ' have values for calculated stress and ratio of calculated stress to allovable stress that were good at one time but may change over the life of the plant. The calculated stresses are supplied for information only and represent actual stresses calculated at a point in time. . Summary *
-I. No. .This is not a change to the plant and the components described still !
meet the ASME design requirements described in the USAR. ! II. No. This is not a change to the plant and is only a change to how calculated stresses are listed in the USAR. III. No. The margin of safety required by ASME is maintained. i' SE No.: 88-508 Source Document: DCp 88-0048, Rev. O I Description of Change { This design change removes access platforms for valves IB33-F060A/B to facilitate disassembly of these valves for required maintenance. - Summary ' J I. No.- The referenced platforms were installed for personnel access only, to accommodate service of valves 1833-F060A/B. The platforms do not , provide any other function. Therefore, their removal,-to provide [ adequate clearance for valve maintenance, does not increase the l probability of occurrence or the consequences of an i accident or malfunction of equipment important to safety. II. No. Removal of_the platforms does not create the possibility for an accident / malfunction of a different type than previously evaluated
-in the USAR.
III. No. The platforms are not described in any Technical Specifications. Accordingly, the margin of safety as defined in the bases of any i Technical Specifications is unchanged. f
n l
' Attcchment 3 PY-CEI/NRR-1141 L Page 95 of 285 SE No.: 88-509 Source Document: DCP 88-0338, Rev. O Description of Change ,
This design change removes vent valve IP11-F0602 from the Condensate Transfer and Storage (P11) System and replaces the vent with a velded pipe cap. This DCP also replaces vent valve IP11-F0575 (Kerotest) with a Marsh needle valve. Summary I. No. The removal of vent valve IP11-F0602 and the replacement of vent valve IP11-F0575 (Kerotest) with a Marsh needle valve on the 10-inch diameter return header from the RCIC and HPCS pumps used during testing, and from the control rod drive pumps used during minimum recirculation, does not increase the probability of an accident or malfunction of equipment important to safety as described in Sections 9.2.6 and 9.2.6.2 of the USAR. Section 9.2.6 describes the Condensate Storage Facilities. Valves IP11-F0602 and 1P11-F0575 are not addressed in the USAR: however, the above mentioned header is addressed in section 9.2.6.2. The valve removal or replacement vill not change the original function of the header. Therefore, the vent valve removal or replacement does not decrease safety in any way, as described in USAR sections 3.5.1.4, 6.3.2.2.1 and 7.4.1.1. This DCP does not affect any of the design basis accidents as described in USAR Chapter 15. II. No. The vent valve removal and replacement vill not affect the original function of the 10-inch diameter return header described in Item I of this evaluation and thus vill not increase the probability for an accident or malfunction of a different type than any evaluated previously in the USAR. III. No. The vent valve removal and replacement does not reduce the margins of safety as defined in Technical Specifications Sections 3.5.2.e.2, 4.5.2.2 and 3.5.3.b.3 which address the Condensate Storage Tank minimum volume of water to support the ECCS. l l
j '. 'Attcchment 3 PY-CEI/NRR-1141 L Page 96 of 285 SE No.: 88-510 Source Document: PAP-0507, Rev. 8 h Description of Change Administrative instructions for the Administrative,. Cost and Programming, Licensing and Compliance, and Operation Services Sections are being 1 combined into one volume of the Operations Manual. Additionally, two p volumes (7D and 7E) are being added to Volume 7 of the Operations Manual for the Ten-Year In-service Inspection Program Plari and the Pump and Valve In-service Testing Program Plan. Summary I. No. This administrative change does not impact plant safety. The combining of administrative instructions into a single volume does not change the review or approval requirements for the documents. The inclusion of the ISI Program Plans into the Operations Manual provides added control of the review and approval of these documents. II. No. This administrative change does not create the possibility for an accident or-malfunction. III. No. This administrative change complies with Section 6 of the Technical Specifications.
i Attcchment 3 ! PY-CEI/NRR-Il41 L I Page 97 of 285
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SE No.: 89-001 2 Source Document: DCp 88-0372, Rev. O Description of Change F 1 i Modify the " Main Turbine'& Feedpump Trip-L8" annunciator logic by adding L a Reactor. Core Isolation Cooling (RCIC-E51) System initiation contact in J { parallel =vith-the existing logic. The new annunciator vindow vill be , worded " Main Turbine and FDW Trip RCIC/L8". This design change corrects : Human Engineering Deficiency 615 and prevents operator confusion and ! delay in emergency response because of the inability to directly determine that the Reactor Feed Pump Turbines (RFPTs) had tripped due to ! RCIC initiation. ' F Summary
- 1. No. This modification only impacts the annunciator logic, it does not ;
impact the automatic actuation circuits nor impact the operator actions required to mitigate-the Chapter 15 accidents. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAr,is not increased. II. No. This modificati.on to the annunciator logic does not impact equipment needed for safe shutdown. Therefore, the possibility of creating a new kind of accident does not exist. 111. No. Modification to this annunciator logic does not impact any Technical Specification basis therefore, the margin of safety has not been decreased.
Attcchment 3 PY-CEI/NRR-1141 L Page 98 of 285 SE No.: 89-002 Source Document: SOI N27, Rev. 6, TCN 8 ;
-Description of Change I Addition of Section 7.17. Pumpdovn of the Hot Surge Tank, to operating instruction SOI-N27. This section uses the B Reactor Feed Booster Pump '
(RFBP) with the low suction pressure switch jumpered out to pump the Hot Surge Tank to the Hotvell prior to maintenance activities in or on the tank. Summary I. No. Neither the probability of occurrence nor the consequences of an accident or malfunction previously evaluated in the USAR are increased when operating the RFBP to drain the Hot Surge Tank with the lov suction pressure switch jumpered out. This operation is bounded by the accident analysis in the USAR, Chapter 15, Sections 15.2.7 and 15.6.6. The draining operation is conducted with the system isolated from the reactor vessel and in long cycle cleanup resulting in system parameters and plant conditions that are well withi% those analyzed in Chapter 15. II. No. Operation of the RFBP to drain the Hot Surge Tank with the lov suction pressure switch jumpered out cannot cause an ; accident / malfunction of a different type than any previously evaluated in the USAR. The operation is bounded by the accident analysis in the USAR, Chapter 15, sections 15.2.7 and 15.6.6. The Condensate and Feedvater System are isolated from the primary coolant boundary during this evolution and thus cannot affect any reactor parameters, including temperature, pressure, or reactivity. III. No. No margins of safety as defined in the Bases of Technical Specifications are reduced by operation of the RFBP to drain the Hot Surge Tank with the lov suction pressure switch jumpered out. Neither feedvater flow nor Hot Surge Tank level are used to determine any margin of safety in the Technical Specifications. l b l l l l
Attcchment 3
; PY-CEI/NRR-1141 L Page 99 of 285 SE No.: 89-003 t source Document: DCP 87-0633B, Rev. 0 :
r' Description of Change This design change provides power and control circuitry for the two nev , motor-operated containment isolation valves 1E12-F609 and 1E12-F610. l These valves are installed as part of a design change to provide for suppression pool cooling / cleanup without making a loop of the Residual. l 6l = Heat Removal (RHR) System inoperable. Operating License Amendment 21 < modified plant Technical Specifications before the implementation of this DCP. (Fire Protection Evaluation) Summary
, I. No. This design change installs cables and controls associated with E12-F609 and E12-F610. These valves will serve as an isolation ,
barrier between the Division 1 Residual Heat Removal (RHR) and l Suppression Pool Cleanup (SPCU) Systems. The spurious opening of both these valves can result in SPCU System ilov being added to the Division 1 RHR suppression pool return line. This will decrease RHR System flow and consequently decrease Division 1 suppression pool cooling. Therefore, it is necessary to , show that the Division 1 RHR/SPCU System interface remains closed ' vhenever it is necessary to utilize Division 1 Suppression Pool : Cooling (NOTE: Division 1 Suppression Pool Cooling is required to support Method A shutdown. Method A is the approach utilized to , achieve safe shutdown during a fire utilizing predominantly ! Division 1 equipment). The circuits required to ensure the
-operability of valve E12-F609 have been identified and their routing
- determined (valve E12-F609 is capable of isolating the RHR/SPCU System interface).
For all plant areas except the main Control Room, the cables ' associated with valve E12-F609 are routed in portions of the plant where Division 1 Suppression Pool Cooling or any other Method A functions are not required (NOTE: Hethod B is relied upon to achieve safe shutdown in these areas. Method B is the approach utilized to achieve safe shutdovn during a fire primarily utilizing Division 2 equipment). Therefore, the potential spurious operation of E12-F609 is not a concern in these areas. In addition, since the circuits of E12-F609 are routed only where Method B is the relied upon shutdown approach (excluding the main Control Room) the potential for fire induced faults in these circuits causing the tripping of upstream breakers (and the consequential loss of other Method A loads) is not a concern. l t l' , I
r
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Attcchment 3 PY-CEl/NRR-ll41 L Page 100 of 285, SE No.: '89-003 (Continued) [ Summary For a Control Room fire, Method A is relied upon to achieve safe shutdown. A Control Room fire may cause the spurious operation of E12-F609. This is not a concern in that transfer / control switches for valve E12-F609 have been provided on the remote shutdown panel. These switches are capable of transferring control, isolating Control Room signals.and transferring control power to an alternate I' supply during a Control Room fire. Fire induced faults in these j circuits vill not jeopardire Method A power supplies in that only control. circuits are located in the Control Room. These circuits i are protected by a 5 A fuse. This. fuse vill limit the fault current such that upstream Appendix R power supplies vill not be i jeopardized. . Valve E12-F610 is not required to operate in order to support either f Method A or B shutdown. However, since E12-F610 is powered in - MCC-ErlC07B and that MCC-EFIC07B feeds multiple Method B loads, it is necessary to demonstrate that a fault in the circuitry associated with this valve vill not affeet MCC-EFIC07B's main supply breaker. I According to calculation 686-85-1, a 30 amp fuse vill coordinate with MCC-EFIC07B's supply breaker. Therefore, the 5 amp fuse which > powers E12-F610 vill also coordinate. I Therefore, the probability of occurrence or the consequences of an , accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. ' II. No. This change does not adversely affect the Fire Protection Program, therefore, the possibility of an accident or malfunction of a different type than evaluated in the USAR is not created. , III. No. The only aspect of the Fire Protection Program addressed in the ' bases section of the Technical Specifications is remote shutdown : capability. This DCP adds valve E12-F609 and E12-F610. It is necessary to ensure the ability to close one of these valves whenever Division 1 suppression pool cooling is required. For a Control Room fire, the ability to close E12-F609 has been ensured by placing E12-F609 on the remote shutdovn panel. Therefore, remote shutdown capability has been assured and the applicable bases has ; not been changed. l
Att:chment 3 PY-CEI/NRR-ll41 L Page 101 of 285 SE No.: 89-004 Source Document DCP 87-0633B, Rev. O Description of Change This design change provides power and control circuitry for the two nev motor operated containment isolation valves IE12-F609 and IE12-F610. These valves are installed as part of a design change to provide for suppression pool cooling / cleanup vithout making a loop of the Residual Heat Removal (RHR) System inoperable. Operating License Amendment 21 modified plant Technical Specifications before the implementation of the DCP. (Electrical Evaluation) Summary I. No. This proposed change adds two new outside containment automatic isolation valves 1E12-F609 and IE12-F610 in series with an existing stop check valve. The two new valves vill each have separate power supply via Division I and Division II. The consequence of one of the containment valves falling to close on RPV hevel 2 does not prevent LPCI "A" loop or suppression pool cooling "A" loop from performing its intended design function. Containment integrity is maintained with one of the isolation valves operable. The MOV's, relays and control switches added in the DCP are classified IE and meet the criteria of IEEE 323-1974, as well as the requirements specified in USAR Section 3.11. Therefore the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. See Item 1 above. III. No. The two new valves are installed in series, each having an independent power supply with one valve (IE12-P609) having the capability of being operated from the remote shutdown panel
,(Technical Specification Table 3.3.7.4-1). The proposed modification is therefore consistent with the Technical Specifications concerning valves isolating other penetrations, hence the margin of-safety is not reduced.
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Attcchment 3 l PY-CEI/NRR-1141 L ! ! Page 102 of 285 l S,E No.: 89-005
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Source Document: DCP 88-0260, Rev. O i l ! Description of Change This design change adds a " normal / disable a switch at the motor control , center to the inboard Shutdown Cooling Suction Isolation Valve IE12-F009 { to prevent the possibility of both Residual Heat Removal System suction. Isolation valves opening due to a fire induced short circuit. These != changes were initiated by CE FDDR KL1-6286 because of the Appendix R requirements to keep 1E12-F008 open during normal operation. After l implementation of this DCP, valve 1E12-F006 vill be closed (breaker vill be racked in) during normal operations therefore, the valve IE12-F008 design is restored to the original design. This change also modifies Shutdown Isolation Cooling Valve 1E12-F009, control / alarm circuitry to meet Appendix R requirements as well as human factor concerns. (Fire Protection Evaluation) Summary
- 3. No. The Appendix R analysis requires that the Residual Heat l
Removal / Reactor Pressure Vessel (RHR/RPV) isolation boundary (composed of valve E12-F008 and E12-F009) remain isolated whenever the RPV pressure is in excess of 135 psig. The original analysis accomplished this by racking out the motive power from valve IE12-F008, such that it cannot open spuriously. thereby ensuring RHR System integrity. This DCP modifies the control circuitry of'E12-F009 such that it is not subject to spurious failures and vill therefore remain closed except for a fire in Fire Area 1CC-3A. In this area 1E12-F008 is unaffected. This DCP does not affect the basis of the original Appendix R study in that the RHR/RPV boundary vill remain closed as required. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. This change does not adversely affect the Fire Protection Program, { therefore, the possibility of an accident or malfunction or a different type than evaluated in the USAR is not created. 1 III. No. The only aspect of the Fire Protection Program addressed in the bases section of the Technical Specification is alternate shutdown capability. This change does not adversely affect alternate shutdown capability, therefore, the margin for safety defined in the bases section of the Technical Specifications is not decreased.
a Attechment 3 . PY-CEI/NRR-ll41 L : Page 103 of 285 ; SE No.:- 89-006 I source Document: DCP 88-0260, Rev. 0 : 1 Description of Change See description for Safety Evaluation 89-005. (Electrical Evaluation) i Summary- ! I. No. The ability of the Shutdown Cooling Suction Valves IE12-F009 and i lE12-F008 to close in response to a containment isolation is not - affected because the valve closing control circuits are not changed via this DCP.
'The bypass switch vill prevent IE12-F009 from spurious opening, but retains power to Control Room indicators. The switch is highly
- reliable, qualified for IE service and all additional valve control ,
virings are located within Class IE motor control center-compartment. An addition of inop/ bypass vindow and annunciator will ; alert the operator of an inadvertent use of the switch.
- The consequences of an accident or malfunction causing IE12-F009 to open are not increased. The accident analysis of the USAR Section 5.2.9 remains bounding as that analysis assumes that the operator is unable to open the suction isolation valves by any means.
- 11. No. This change affects the opening circuit of the IE12-F009. As the complete failure of this valve to open is assumed in the USAR >
Section 5.2.9, no additional malfunction or accident is created. As the valve closing circuits are unaffected, no additional concerns ' relating to containment isolation or Reactor Coolant Pressure (RCP) boundary integrity are created. III. No. The shutdown cooling function of the Residual Heat Removal System is not impaired by this change. Therefore, the redundancy of shutdown cooling methods assumed in the basis for Technical Specification 3/4.4.9 is not reduced.
p.- _- h Attcchment 3 4 PY-CEI/NRR-1141 L [ Page 104 of 285 t i SE No.: 89-007 ' Source Document: SXI-0034 i Description of Change I This instruction performs an operational test of the Radiation ! Protection Data Information System program. This is a one-time ; L software verification test of an administrative. computer program used to track radiation exposure. The test of this program is to verify that the system meets the requirements in USAR Chapter 12,- ; Sections 12.5.3.6 and 12.5.2.3.3. l
; Summary i
i 1. No. This is an administrative computer program and does not interact or affect any safety-related systems or components. e II. No. This is an administrative computer program and does not interact ; with any of-the systems or components listed in USAR Table 15.0.3. - It does not create the possibility for any new type of accident. t III. No. Since this computer program does not interact with any plant systems .i or components, the margin of safety in any Technical Specification ; is not affected, i
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Attcchment 3 7 PY-CEI/NRR-1141 L L Page 105 of 285 , SE No.: 89-008 Source Document SXI-0035 i l Description of Change This instruction describes preoperational testing of the P85 Iontrac 200 , Zine Analyzers 1P85-D001 and IP85-D002. Summary I
- 1. No. Preoperational testing of the P85 Iontrac 200 Zine Analyzers IP85-D001 and IP85-D002 in no way alters the function or operation of the Reactor Vater Cleanup System (RVCU G33) System, Nuclear Closed Cooling (P43) System, or the Turbine Plant Sampling (P33) i System as described in USAR sections 5.4.8, 9.2.8 and 9.3.2 respectively. These analyzers, which monitor zine levels in the reactor water and feedvater systems, sample continuous streams provided from the G33 System (i.e., RVCU pumps) and the P33 System. >
The G33 flovpath utilized a RVCU pump sample recirculation loop connected to the existing vent and drain valves of each pump. This - line was designed, installed, and tested under the same ANSI B31.1 ; requirements as the existing RVCU piping. Further, the introduction
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of this sample flovpath limited the worst case scenario to a sample ,. line break. This scenario is bounded by both G33 and E12 break analyses described in the USAR. The P33 flovpath branches off an existing feedvater sample line at the turbine plant sample panel. , The P43 System merely cools the reactor water sample stream prior to entry into the analyzer. II. No. This test changes the design configuration of the RVCU System as described in the USAR. The configuration change involves altering the valve positions for the existing RVCU pump (1G33-C001A/B) vent ' and drain valves from normally closed to normally open to accommodate the PBS sample line. The vorst case scenario for this sample flovpath is a line break. This is described in Item I above. ' Therefore, no possibility exists for an accident different from that previously evaluated in the USAR. 111. No. Since the likelihood of an occurience or the consequences of an accident to either G33, P43, or P33 resulting from P85 system operation do not increase, the function of these systems remain ' unaffected. As a result, the margin of safety as defined in the Technical Specification bases remain unchanged.
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'Attcchment 3- j PY-CEI/NRR-1141 L Page 106 of 285 ;
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... SE No.: 89-009 l
[- Source Document: DCP 86-0005, Rev. O
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3-Description of Change i
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This design change replaces the Two Bed Demineralizer (P21) System j Recycle Rinse Cycle Valves OP21-F010A/B vith solenoid valves which vill l[ ~
-fail close and energize to open. ,
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- Summary ~
- 1. No. Solenoid valves OP21-F0010A/B are used for demineralized water I sampling and their changed failure position does not alter the function or operation of the P21 System. Therefore the probability '
{ of occurrence or the consequences of an accident or malfunction of . equipment'important to safety previously evaluated is not increased. ; r II. No. See Item I above. III. No. Since those valves do not alter the function or operation of the p P21 System nor are they listed within the Technical Specifications, +
.the safety margin is not reduced.
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Attcchment 3
- PY-CEI/NRR-1141 L Page 107 of 285 ;
SE No.t 89-010 ' source Document: DCP 88-0305, Rev. 0 , Description of Change The current.pover response rate for Perry (1% pc. second, 60% per minute) L is unrealistically high. This design change vill reduce the power response rate to a more standard 1% per minute by adding rate-limiting devices to the Feedvater control (C34) System and retuning the Reactor Recirculation control (B33) System. Summary
- 1. No. The Perry Nuclear Power Plant has been tuned for a power response rate of 60% per minute. Expetience with large power plant installations has shown that this is an unnecessarily rapid response and that actual rates of 1% per minute are more standard. This design change reduces the reedvater System response (by rate-limiting) and retuning the Reactor Recirculation System to a slover response which will enhance overall control system '
operability. These changes are based on an analysis performed by General Electric (EDE-31-0988). The analysis was performed with the assistance of the REDYV06 computer program, a digital computer program which simulates the dynamic behavior of boiling vater reactors. This - program was used to develop the original design transient analysis discussed in the Perry Control System Reports (GEZ-7134). This DCP will allow the systems to meet the same macro requirements as before. The macro requirements which must be met are as follows:
- a. Avoidance of a lov vater level scram (Level 3) due to a feedvater pump trip.
- b. Avoidance of Level 2 and Level 8 trips due to main generator or turbine trip.
- c. Maintain the systems stability.
- d. Maintain adequate margin to scram for other design basis events (DBE) transients for which a scram is required to assure safe rhutdown.
llence , the Reactor Recirculation and Feedvater Control Systems vill operate to the original design intent. The reduced plant power response vill not affeet plant reliability or any safety systems. ' Therefore the probability of occurrence or consequence of an + accident or malfunction of equipment to safety previously evaluated is not increased. Since plant operation vill be enhanced and the original design intent of the Reactor Recirculation and Feedvater Control Systems have not been changed, no new failure modes have been created.
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Att:chment 3 i PY-CE1/NRR-1141 L i L Page 108 of 285 i SE No.: 010 (Continued) > Summary i: ; II. No See Item I above. l III. No.; This change does not affect the Technical Specifications thus the ' c margin of safety has not been reduced. , I SE No.: 89-011 Source Document: DCP 87-0196, Rev. 0 - i Description of Change ' Expand the Maintenance and Calibration Communication (R52) System in the [ Unit 1 Control Room by installing jacks in 'the 1H13-P870J and 1H13-P877A ! panels.' L Summary F e
- 1. No. This R52 system expansion has no interconnect or involvement'vith ;
L any plant process or safety systems. ~ II. No. This change vill reduce the possibility for an accident by l eliminating the need to stretch an RS2 headset cord across the ! p entire front of the ECCS and BOP control room bench boards which ! reduces the chances of a switch entanglement or personnel tripping : l mishap in the " horseshoe area of the control panel". !
'III. No. This change does not reduce or affect any Technical Specification' I margin of safety (see 3/4.9.5 - Refuel Communications).
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L Attcchment 3 PY-CEI/NRR-1141 L Page 109 of 285 , SE No.: 89-012 Source Document: DCP 86-0748A, Rev. 0 l Description of Change The Reactor Vater Clean-up System Outlet Valve, 1G33-F042, and the RVCU/ Filter-Demineralizer Bypass Valve, 1G33-F044, are being replaced with motor operated globe valves with improved throttling ' characteristics. Further remote valve position indicators (percentage of valve opening) 1G33-R0612 and 1G33-R0613 respectively on_the plant Unit Control Console (1H13-P680) for these two throttle valves has been added. Summary e I . No. - This design change concerns the addition of control position indication and valves exhibiting better throttling characteristics ' for the Reactor Vater Cleanup (RVCU) System valves 1G33-F042 and : 1G33-F044. These valves are designated nonsafety and are not described or taken credit for in any USAR safety analysis in i Chapters 5, 7, or 15. The addition of position indication circuitry ; and components is being designed nonsafety/non-1E, consistent with this portion of the G33 system design. It is concluded that the probability of occurrence or the consequences of an accident / i malfunction of safety-related equipment previously evaluated in the ' USAR is not increased. 1 II. No. The scope of this design change is limited to the modifications described above. No other systems or equipment are affected by the , t subject DCP. A different type of accident or malfunction than previously described in the USAR is thus not created. , III. No. Technical Specifications and associated bases do not address design or operability requirements for RVCU system valves 1G33-F042 or 1G33-F044. Addition of Contiol Room position indication for these valves vill therefore have no impact on the Technical Specifications. The margin of safety defined in any Technical J Specification basis is not reduced. L 5 I e
l' L' h Attcchment 3 J PY-CEI/NRR-Il41 L ; Page 110 of 285 SE No.: 89-013 ! Source Document: DCP 85-2951, Rev. 1 j Description of Change i.. This design change adds a duct-mounted smoke detector in the exhaust [ plenum of the Small Tool Decon HVAC Charcoal Filter Unit located at the i
! .. 574' Elevation of the-Intermediate Building. ;
E Summary [ I. No. This change adds a type of fire detection (duct-mounted smoke [ detector) to Fire Area IB-1 that is different than the existing (' detection in the area. This type of detector is required for the
. fire hazard introduced by the combustible charcoal in the HVAC filter. The detecto'r is installed in accordance with the , requirements of NFPA 72D; Installation, Maintenance and Use of ! Proprietary Protection Signaling Systems; and 72E; Automatic Fire Detectors, as described in the USAR. This detector will provide the same level of protection as previously evaluated in the USAR for this type of fire hazard. ; II. No. Although this change introduces additional fire protection, the type of hazard being protected is additional combustibles. Protection of these combustibles is within the scope of the Fire Protection Program; therefore, the possibility of an accident or malfunction of.
a different type than evaluated in the USAR is not created. o. L- III. No. Only administrative aspects of the Fire Protection Program and Alternate Shutdown capability are defined in the Technical Specifications. This change has no impact on the administrative aspects of-the Fire Protection Program and does not adversely affect Alternate Shutdown Capability. Therefore, the margin of safety as defined in the basis for the portions of the Technical Specifications applicable-to fire protection vill not be affected. l 1 -- l lI
Attachment 3 PY-CEI/NRR-1141 L Page ill of 285 SE No.: 89-014 Source Document: PAP-1915, Rev. 2 Description of Change This safety evaluation evaluates thanges to procedure PAP-1915, Fire Report. The changes made were administrative only and are consistent with the Fire Protection Program as specified in the USAR. Summary I. No. All changes made to this procedure were administrative in nature. All changes were found to be consistent with the fire protection requirements of the USAR and the Fire Protection Program. Therefore, the probability of an occurrence or consequence of an accident previously evaluated in the USAR was not increased. II. 'v. See Item I above. III. No. The Fire Protection Program is referenced in Sections 6.5.1.6, 6.5.2.8, and 6.8.1 of the Technical Specifications. Those changes are consistent with the Fire Protection Program, hence are consistent with the Technical Specifications. Therefore, no margin of safety has been reduced.
Attachment 3 PY-CEI/NRR-1141 1. Page 112 of 285 SE No.: 89-015 Source Document: PAP 1910, Rev. 2, TCN 1 Description of Change Evaluation of a Temporary Change Notice to PAP-1910, Fire Protection Program, that modifies a note !n Section 6.7.1 concerning Fire Brigade staffing. This note is being revised to read, "The Shift Supervisor shall not be assigned to the Fire Brigade. The Operator 'with the book' should-not be normally assigned as the Fire Brigade Leader. Additionally, assignment of Operations personnel fulfilling the required Minimum Shift'Crev should be minimized". Summary See Safety Evaluation 88-470. SE No.: 89-016 Source Document PNPP Emergency Plan, Rev. 9 Description of Change This evaluation analyzes varicus changes made to this revision of the
'PNPP Emergency Plan to ensure that the effectiveness of the plan has not been reduced, per 10 CFR 50.54(q), and to ensure that the plan continues to meet the standards of 10 CFR 50-47(b) and the requirements of 10 CPR 50 Appendix E.
Some of these changes involve updated references to State, Local and CEI titles to reflect current organizations and training requirements, clarified or added Emergency Action Level (EAL) indications, and updated plant specific technical / system information. These changes also reflect Lake County Emergency Management Agency comments,-and clarify requirements for unannounced exercise and testing of the Prompt Alert System. Summary I. No. The Emergency Plan outlines administrative (cmergency preparedness) response to an accident and therefore does not affect the probability of occurrence or the consequences of an accident or malfunction of equipment. II. No. The Emergency Plan does not direct the operation of plant systems or equipment and therefore, does not create the possibility for an accident or malfunction. III. No. The Emergency Plan utilizes existing Technical Specifications. Therefore, the margin of safety as defined in Technical Specifications is not reduced.
Attachment 3-PY-CEI/NRR-1141 L Page 113 of 285
.SE No. 89-017 Source Document PAP 0101, Rev. 4 Description of Change This evaluation reviews changes made to PAP-0101, Perry Plant Operations /
Technical / Nuclear Support Department Organizations, concerning the responsibilities for maintenance and operation of the Control Room Simulator and for evaluation of plant thermal-hydraulic response, l Summary I. No. The probability of an accident or malfunction vill not be affected because the responsibilities for performing the tasks are not deleted, only reassigned. II. No. The poss.ibility of accidents or malfunctions different than those i evaluated will not change because the tasks being reassigned vill continue to be performed. III. No. This administrative document is in compliance with the administrative controls required by the Technica'l Specifications.
Attachment 3 PY-CEI/NRR-1141 L Page 114 of 285 SE No.: 89-018 Source Document: DCN 02544, Rev. O Description of Change This drawing change involves correcting the MPL labeling on the Condenser Air Remover (N62) System B loop intercondenstr loop seal instrument root valves on P & ID 302-131. Summary I. No. This DCN simply corrects the labeling of the intercondenser loop seal level instrument root valves on P & ID 302-131. It does not change the function of the N62 System. Therefore, this DCN vill not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as defined in the USAR. II. No. See Item i above. III. No. The Condenser Air Removal System still retains its original function with the cortection of the MPL labels on P & ID 302-131. This DCN vill not affect the operability or the availability of the Offgas System (Base 3/4.11.2.4) or the Ventilation Exhaust Treatment System (Base 3/4.11.2.5). The hydrogen or any other explosive gas mixtures concentration vill not be increased (Base 3/4.11.2.7). Therefore this DCN vill not reduce a margin of safety as defined in the Technical Specifications.
Attachment 3 PY-CEI/NRR-1141 L Page 115 of 285 SE No.: 89-019-Source Document: DCP 88-0234, Rev. O Description of Change This design change adds a " fill" line from the Service Water (P41) System to the Emergency Service Water (P45) System Loop C. This line vill serve as an alternate means of filling Loop C prior to plant startup. The line. t consists of 2-inch diameter piping, 3 inch diameter safety-related piping, and a 3-inch diameter safety-related, normally closed gate valve. This line vill be run parallel to the existing 3/4 inch keep fill line to [^ Loop C. e Summary I. No. The Emergency Service Vater System (ESV) is classified as Safety Class 3 and Seismic Category I, with the primary safety function being to support the Emergency Core Cooling System. The 2-inch
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diameter fill line vill decrease the time required to fill Loop C of the ESV. This change does not alter the original function of ESV Loop C or the 3/4-inch keep fill line to Loop C as described in USAR Sections 9.2.1.1 and 9.2.1.2. The nonsafety portions of the 2-inch diameter quick fill line and the 3/4-inch diameter keep fill line vere leak tested per ANSI B31.1. The affected 3-inch diameter piping was tested in accordance with ASME Section XI, Article 1 VA-500. Therefore the alternate fill line addition vill not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. II. No. See Item I above. III.-No. The piping modifications covered in this DCP do not reduce the margin of safety as defined in the Technical Specification Sections: Tables T3.3.7.7-1 Remote Shutdown System Instrumentation and T3.3.7.4-1 Radioactive Liquid Monitoring Instrumentation of ESV Loops A & B, 3.7.1.1 ESV Limiting Conditions of Operation; and 3.8.1.1, ESV Electric Power Systems. E E L . . ._ . _ . - . _ _ . _ . _ . - _ . _ _ _ _
Attachment 3 PY-CEI/NRR-1141 L Page 116 of 285 SE No.: 89-020 Source Document: USAR CR 89-019 Description of Change This change request revises USAR Section 1.7.1 to clarify which drawings are included in the USAR, and Table 1.7-1 to indicate the latest revision of those drawings. This change also formally replaces the GE Functional Control Diagrams with the corresponding CEI 808 Series drawings. Summary I. No. The individual changes contained in this change request have already been safety evaluated and approved. This is simply a compilation of those changes. Safety Evaluations 88-270, 88-271, 88-272, 88-273, 88-274, 88-275, 88-276. 88-277, 88-282, 88-284 and 88-286 provide justification for the addition of the 808 series drawing. Safety evaluations for the individual drawing changes identified in Table 1.7-1 provide justification for changes to the revision levels in Table 1.7-1 and for the discussion in Section 1.7.1. II. No. This change provides no new information which could affect the plant or equipment. It is simply a compilation of other changes. III. No. .The changes contained herein have already been individually approved. This update to the USAR figure listings does not affect the Technical Specification bases. (
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!Attachm2nt'3 PY-CEI/NRR-1141 L Page 117 of 285 SE No.: 89-021 Source Document: -USAR CR 89-021 ,
Description of Change ThisiUSAR change request-corrects various editorial errors in the original issuance of the USAR. l Summary I. No.. Changes are editorial in nature, e.g., spelling, grammar, format, etc. No accident or malfunction to existing equipment is involved. s II. No. Changes are editorial. No new accident or malfunction is involved. III. No. No Technical Specification margin of safety is involved. SE-No.: 89-022 Source Document: USAR CR 89-022 Description of Change This USAR change request corrects partially incorporated change requests and makes various other editorial corrections. Summary I. No. These changes have already been approved either in the FSAR or in safety evaluations previously approved for the USAR update,~but were not incorporated into the USAR correctly.
.II. No. These changes are already approved and create no new accident or malfunction.
III.'No. These changes are already approved and have no effect on Technical Specifications.-
Attachment 3 PY-CEI/NRR-1141 L Page 118 of 285 SE No.: 89-023 Source Document DCP 86-1022B, Rev. O Description of Change This design change adds an instrument air (P52) supply line to provide a breathing air connection and a maintenance air connection and adds a domineralized vater supply and maintenance connection adjacent to the HEPA' Filter Change Room. Drainage piping vill run from a floor drain connection (for a portable sink) inside the HEPA Filter Change Room to the existing (radioactive vaste) Floor Drain Collecting System. Summary I. No. The instrument air piping addition does not increase the probability of an accident or malfunction of equipment important to safety as defined in USAR Section 9.3.1. The Instrument Air System has no safety-related functions as defined in Section 3.2. Safety-related devices. supplied with compressed air from this system are designed
'for the fail-safe mode and do not require continuous air supply-under emergency or abnormal conditions. The added air piping contains a check valve to prevent contaminated air from entering into the system.
USAR Section 15.2.10 describes the results of a major line break in the Instrument Air System. The 1-inch line added by this DCP is not classified as a major line. The major line break has been previously evaluated and determined not to interfere with safe shutdown. USAR Section 9.2.3 addresses the Two-Bed Demineralized Vater Makeup (P21)-System. The system is nonsafety-related. USAR Section 9.9.3 addresses the Equipment and Floor Drainage System (EFDS). The EFDS is designed to provide for collecting radioactive and potentially radioactive liquid vastes from the floor drains and equipment drains throughout the plant, and convey these vastes to building sumps located in the-basements of the major structures. This DCP adds a hard piped drain from a portable sink inside the HEPA Filter Change Room to an existing floor drain outside the room. Therefore, these additions vill not increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR. II. No. See Item I above. III. No. This design change does not impact the basis of any Technical Specification.
v Attech:3nt 3 PY-CEI/NRR-1141 L I Page 119 of 285 No.: 89-024' Source Document: DCP 88-0262, Rev. O Description of Change Installation of the necessary piping / tubing, isolation valve, orifice and-supports to allow relication of the "Between the Seals" test connection from the dryvell inner-airlock door bulkhead to the dryvell outer airlock i door bulkhead. This will reduce dryvell outer airlock door operation and , reduce personnel radiation exposure when testing is performed at power. Summary I. No. Chapter.6.2 Cor.iainment Systems, of the USAR discusses justification with respect to General Design Criteria (GDC) 56 of 10 CFR 50. Although GDC 56 is concerned with containment, the criteria is conservatively applied to dryvell penetrations to minimize potential ' bypass leakage. GDC 56 allows deviation from the required two valve arrangement for isolation if "it can be demonstrated that the containment isolation provisions for a specific class of lines,'such as instrument lines, are acceptable on some other defined basis." The design provided by this modification is consistent with American National Standards (ANS) 56.2/N271-1976 Containment Isolation Provisions for Fluid Systems. Although installation of this'line creates a potential path for bypass leakage, Section 6.2.1.1.5 of the USAR discusses this topic in terms of total leakage. Total leakage vill not significantly increase singe the 1/4-inch orifice limitsthepotentialleakpath-to0.00g4ft . The Technical Specifications for leakage is 0.168 ft . Further, the-design for this line requires ASME Section III Class 2 materials / components with Seismic Category I support. The line is normally depressurized with'the manual valve closed and the test l connection capped. It is used for between the seals. testing of the I inner dryvell personnel airlock door. Testing is in accordance with an approved surveillance test instruction. l Therefore, the probability of occurrence or the consequences of an-l
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accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II No . - As described in Item I above, the design requires ASME Section III Class 2 requirements be met. Further, no active components are installed or affected by this design change. Hence, the possibility ' for an accident or a malfunction previously evaluated is not created. III. No. This design creates a potential leakage path of .0004 ft 2 This is l a small fraction (.2%) of a leakage allovable by Technical Specification. Further, Technical Specifications is based upon the total leakage not upon leakage from any one point. Therefore, the margin of safety is not reduced. l l 1
Attachmsnt 3. PY-CEI/NRR-1141 L-Page 120 of 285' SELNo. = 89-025 Source Document: DCP 87-0800, Rev. 0-
. Description of Change Replace the existing 1-1/2 inch turbine flovmeter (ITT Barton) in the Leak Detection (E31) System with two 1 inch magnetic flovmeters (Rosemount). The turbine meter measures identified leakage from the dryvell air coolers and has had a history of problems with clogging.
Summary I. No. Flov elements 1E31-N021, and 1E31-N022, and the associated transmitter 1E31-N695 are nonsafety-related, Seismic Category I. The system for monitoring the Reactor Coolant Pressure Boundary leakage in the dryvell is not part of the' plant protection system. USAR Sections 5.2.5.2.1 Detection of Leakage within Dryvell; 5.2.5.2.1.c, cooler Condensate Drain; 5.2.5.10 Regulatory Guide 1.45 compliance; and 9.4.6.2.1, Dryvell Cooling System are not affected by this change since only the method of detection has changed (turbine meter versus magnetic flow meter).
-Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased.
II. No.1 Concern was expressed that the conductivity value of the condensate from the dryvell coolers vould not be high enough to support operation of the magnetic flow tubes. The magnetic flov tubes require a minimum conductivity of two microsiemens/cm (2 micromhos/cm or greater). A series of three samples were taken.
/Two from-the Dryvell Cooler drain lines and one from the dryvell atmospheric radiation monitor. Values of 40.5, 40.5 and 17 microsiemens/cm respectively were obtained.
A biological culture (algae / bacteria) has been identified in the condensate drain lines. The vendor has confirmed that there vill be no affect on the function of the flow elements. Technical Assignment File (TAF) 80-297 (Functional Test to determine the effects of biological growth inside the Rosemount Model 8701 magnetic flovtube) was implemented. The results being that no output = signal deviations were noticed during the testing period. Based upon the conductivity values and vendor / testing confirmation, there is no possibility for an accident or malfunction of a different type than any evaluated previously in the USAR. l l
Attachment.3 PY-CEI/NRR-1141 L Page 121 of 285 SE No.: 89-025 (Continued) Summary III. No. The Technical Specification, Section 3.4.3.2, identifies only total allovable leakage. Flow rates, setpoints, alarm functions, instruments / loop accuracy, etc., are not affected. The Reactor Coolant Leak Detection System identified under Section 3.4.4-9, 3.4.3.1, vill remain the same. Therefore, the margin of safety.as defined in the basis for any Technical Specification is not reduced. SE No.: 89-026 Source Document: LL&JED 1-89-161 Description of Change This LL&JED removes temperature switches 1E31-N351A-D from the Main Stream Isolated Valve (MSIV) logic by installing temporary jumpers. They are redundant to the steam. tunnel ambient and delta temperature switches 1E31-N604A-F and N605A-F, located in the same environmental zone. Ilovever,1E31-N351A-D vill still provide input to Control Room alarms. Summary I. No. -Temperature switches 1E31-N351A-D, for the Turbine Power Complex were added by Gilbert Commonwealth to be redundant to the steam tunnel ambient temperature switches 1E31-N605A-F. Operation of the Turbine Power Complex switches, is not assumed in the plant safety analysis, the GA design specification 22A3739AD nor the plant Technical. Specifications. These temperature switches are located in a nonsafety, non-seismic structure. Deleting the Turbine Power Complex temperature switches from-the MSIV isolation logic vill not increase the probability or the consequences of an accident since the Main Steam Line-(MSL) temperature switches N604/N605 (located in the same environmental zone) vill activate the isolation logic if a steam leak occurs. The N604/N605 switches are located in a seismic category I building and credit is taken for them operating, in the safety analysis, GA design specification and Technical Specifications. II..No. The MSL temperature switches vill initiate the MSL isolation-logic on a steam leak. Since the Turbine Power Complex temperature switches, N351 are redundant to-the MSL temperature'svitches, (N604/N605) their deletion vill not cause and accident.
'III. No. The HSL and Turbine Building temperature switches are described Technical Specification 3.4.3.2, Isolation Actuation Instrumentation. The Turbine Power Complex temperature switches are not as described in Item I above, those switches are redundant to N604/N605. Therefore their detection vill not decrease the margin of safety described in Technical Specifications.
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s -Attachment 3 PY-CEf/NRR-1141 L n, m Page 122 of 285 SE No.s c 89-027 Source Document: NR PPDS 3603, Rev. O Description of Change The upper shield block cannot be installed at the Unit 2 Containment Equipment Hatch due to conduit interference. These conduits are permanently installed to support Unit 1 operations. This safety
- evaluation analyzes the affect of not having this shield block installed during Unit I refueling operations.
Summary I. No. The design purpose of the shield blocks is to provide a radiological barrier between the operating Unit 2 and the Fuel Handling Building, and to provide an air seal to ensure proper operation of the Unit 2 Annulus Exhaust Gas Treatment System (AEGTS). Additionally, the eighth block (the block in question) is to prevent shine from an operating unit 2 in the Fuel Handling Building. Since Unit 2 is not operating at this time, the radiological barrier and the eighth block are not required. Also Unit 2 AEGTS is not required to be operational.
-The deletion of the shield block does not affect the ability of the Fuel Handling Building Ventilation System from maintaining a negative pressure in the Fuel Handling Building in the event of a fuel handling accident. Hence no potential from an unmonitored release path to the environment is created.
The deletion of one shield block vill not impair the seismic restraint of these blocks. Therefore, the probability of occurrence or consequences of an accident or malfunction previously evaluated is not increased. II. No. See Item I above. III. No. Technical Specification Section 3/4.7.7.2 requires installation of the shield blocks in order to-establish Fuel-Handling Building integrity. The basis for this is to ensure that there is no potential for untreated release from the Fuel Handling Building. Since the Fuel Handling Building Ventilation System maintain a negative pressure within the building. This ensures that there is no potential for untreated release. Therefore no Technical Specification margin of safety is reduced. I l
. Attachment.3 PY-CEI/NRR-1141 L Page 123 of 285 SE No.: 89-028 Source Document: DCP 87-0793, Rev. O Description of Change This design change reroutes the Combustible Gas Control (M51) backup H2 purge line (within containment) to eliminate water traps inside Containment, replaces Dresser Piccolos globe valves in the purge line with gate valves to reduce line losses and adds a nonsafety-related drain line extension to assist the drainage of condensate from the purge line.
Summary I. No. The piping modifications does not change the operation or function of the M51 backup H2 purge line. All replacement valves and safety piping are consistent with the original safety class and construction codes specified in Table 3.2-1 of the USAR. The reroute of the piping meets all of the seismic, thermal, break exclusion, and ISI requirements of the original design. In addition,-the reroute was analyzed in accordance with USAR Section 3.9 to confirm that the postulated breaks specified in USAR Chapter 3.6 are still valid and that no additional breaks need to be postulated. The nonsafety-related drain line extension is located outside of containment downstream of the containment isolation valves. It is not-part of the system pressure boundary nor does it affect the-backup purge line from performing its intended safety design function.- Its purpose is to allow condensate to be drained from the system directly to the local floor drain. Thus, there is no impact to the previous evaluations in the'USAR as a result of this design change. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. II. No. The M51 Backup 2H Purge System vill function in the same manner as before. Thus the possibility of a malfunction or an accident of a different kind has not been created. III. No. The M51 System vill operate exactly as before. The changes to the system vill only improve its efficiency. Thus, the margin of safety in the Technical Specifications is not reduced. l
Attschmtnt 3 PY-CEI/NRR-1141 L Page 124 of 285 'SE No.: 89-029 Source Document: DCP 88-0317 Rev. O Description of Change Redesign of the Health Physics (HP) and Chemistry Offices on the 599' Elevation of the Control Complex. Specific changes involve enlarging the Chemistry Office by removing a vall, creating a new entrance to the HP Offices, converting the Medical Aid Room into two offices, and divide the Mens Room. This safety evaluation analyzes the affects of the revised airflows required on operation of the Controlled Access Area HVAC (H21) System. (Mechanical-HVAC Evaluation) Summary .t I. No. This design accommodates architectural changes to the controlled access area office space. The function of the M21 System has not ! been altered. Purther, M21 is not required for the safe shutdown of the plant. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated is not increased. II. No. This change does not alter the function of M21. The possibility for an accident or malfunction of a different type than previously evaluated is not created. III. No. The M21 system:is not described in the basis of any Technical Specification. Therefore, no margin of safety has been reduced.
Attachment 3 PY-CEI/NRR-ll41 L Page 125 of 285~ SE No.: 89-030 Source Document: DCP 88-0317, Rev. O Description of Change See Safety Evaluation 89-029 for a general description of this design change. This safety evaluation analyzes the affects of demolishing and capping portions of the Potable Vater (P71) and Floor and Equipment Drains (P68) Systems to support the architectural changes. (Mechanical-Evaluation). Summary I. No. Demolition and capping of the affected portions of these systems is necessary to_ allow for the modifications of the Control Complex outlined under the civil portion of this Design Change. The primary functions of these systems remains unchanged in the areas that vill continue to be-serviced by them. These systems do not affect any safety-related components or equipment. Therefore, the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR has not increased. II..No. The P71 and the P68 Systems do not have a safety-related function per USAR Sections 9.2.4 and 9.3.3, respectively. ' Failure of either of these systems vill not compromise any safety-related system or component and vill not prevent safe reactor shutdown. Hence, the possibility for an incident or malfunction of a different type than previously evaluated is not increased. III. No. The margin of safety described in the bases of any Technical Specification is not decreased since this portion of the Design Change'only addresses the demolition of piping which is not contained in the Technical Specifications.
'Attschrent 3 PY-CEI/NRR-1141 L Page-126 of 285 SE No.: 89-031- ' Source Document: DCP 88-0317, Rev. O O ' Description of Change See Safety Evaluation 89-029 for a general description of this design change. This safety evaluation analyzes the affects of rearrangement of architectural non-load bearing-valls. (Civil / Structural Evaluation). . Summary I. No. The rearrangement of architectural non-load bearing valls and doors vill not affect the structural integrity of the Control Complex Building. The design of the vall along column rov '3' has been' re-evaluated and verifies that the new openings do not impair the-seismic design requirements of the vall. Therefore.this change vill not increase the probability of the occurrence or the consequences of any accident.
II. No. The architectural non-load bearing valls have no safety-related function. The seismic design of the vall-along column rov '3' has been maintained. Therefore this change vill not create any new accident or' malfunction. III. No. This change involves moving architectural non-load bearing valls i which are not addressed in the Technical Specifications, llence, the margin of safety designed into the components / equipment identified in the Technical Specifications is not decreased. i I l i ? i
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s L Attachmtnt 3
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PY-CEI/NRR-1141 L Page 127 of 285 SE No. - 89-032
- Source Document: FTI E13, Rev. 1, TCN 2 Description of Change This safety evaluation reviews the affects of using the Fuel Handling Building (FHB), Fuel Handling Bridge Crane Monorail Holst to move new fuel when no spent fuel is present in the FHB storage pools.
Summary I. No. Use of the monorail hoist is procedurally restricted to movement of new fuel with no spent fuel in the pool and limited to less than or equal to 6 feet above the racks. The monorail hoist has redundant load paths and the same load cell settings as the main hoist per USAR Section 9.1.4.27.3. Handling errors associated with new fuel are of no nuclear safety concern per USAR Section 9.1.4.3. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. II. No. See Item I above.
- III. No. .There is no margin of safety associated with handling new fuel.
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1 4 Attachm:nt 3 PY-CEI/NRR-1141 L Page 128 of 285 SE No.: 89-033 Source Documents. USAR CR 89-046 Description of Change ' t In accordance with the guidance provided in IE Bulletin 80-10, the operation of the Turbine Building Closed Cooling (P44) System as contaminated must be evaluated relative to the limits in 10 CFR 20 and 40 CFR 190. i Summary I. No. Operation of the Turbine' Building Closed Cooling System slightly , contaminated vill not affect the accidents previously evaluated in the USAR. The nuclides which have been identified in the P44 system are .i Xenon-133 and Xenon-135 in concentrations of approximately 3E-6 uCi/ml and 2E-7 uCi/ml, respectively. A potential release point from the P44 System is tube leakage in the P44 heat exchanger to the Service Water System. The P44 System monitoring indicated that there_vas no tube leakage. Further the service water side of the heat exchanger has been sampled and the analysis showed no detectable activity. 1 Potentialireleases to the environment must be evaluated with respect to 10 CFR 20 and Technical Specifications. There are no limits for Maximum Permissible _ Concentrations listed in 10 CFR 20 Appendix B, Table II. The only limits on dissolved or entrained noble gases are found in Technical Specification 3.11.1.1, at 2E-4 uCi/ml total activity. The current concentration in the system is about 1/100th of-this limit. The evaluation of the potential dose to the environment performed in accordance with-Regulatory Guide 1.109, Calculation of Annual Doses to Man'From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, shows that there is no dose from releasing water with dissolved or entrained noble gases.
- II. No. See Item I above.
III. No. The P44 System is not required by Technical Specifications and as such, has no safety margins associated with it.
Attachment 3 PY-CEI/NRR-1141 L Page 129 of 285 SE No.: 89-034 Source Document: DCP 86-1061, Rev. I t Description of Change _This design change removes the vital Non-Class 1E load V-2-C from the 120 VAC Vital Inverters (R14) system and transfers the load to the Technical Support Center Uninterruptible Pover (R15) System. Summary I. No. This design change removes vital nonsafety load V-2-C from the R14 system and transfers it to the R15 System. The R15 System satisfies all'the power supply requirements of the V-2-C load. This load transfer between power systems of similar capabilities does not affect the Security Plan of Section 13.6.1 of the USAR. , The V-2-C bus does not provide power to safety-related equipment. Failure of this equipment would not affect safe shutdown systems therefore, this change does not increase-the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. II. No. All V-2-C loads are nonsafety. Failure of this equipment vould not affect safe shutdown systems. No new accident possibilities are created by this design change. III. No. This bus and its associated equipment are not addressed in the Technical Specifications. Therefore, this design change does not affect the present margin of safety.
r-i , Attachm:nt.3 L. PY-CEI/NRR-1141 L L Page.130 of 285 SE No.: 89-035 Source _ Document: . DCN 2574 Description of Change This change updates a one-line diagram (USAR Figure 8.3-21) by revising the drawing reference sheet number. Summary-I."No. This' Drawing Change Notice (DCN) is'only an administrative update of the drawing. It does not increase the probability of an accident or , malfunction of equipment. II. No. This DCN is an administrative update only'. No possibility of different type of accident or malfunction is created by this change. III. No. The margin of safety as defined-in the bases for any Technical Specification is not reduced by this DCN since it revises the drawing reference only. SE No.: 89-036 Source Document: DCN 2177 Description of Change This change updates a one-line diagram (USAR Figure 8.32-22) by revising j the drawing reference sheet number, a Summary I . No . - This Draving Change Notice (DCN) is only an administrative update of the drawing. It does not increase the probability of an accident or malfunction of equipment. I 4 II. No. This DCN is an administrative update only. No possibility of different type of accident or malfunction is created by this change. III. No. The margin of safety as defined in the bases for any Technical Specification is not reduced by this DCN since it revises the l drawing reference only. t
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Attachment 3 PY-CEI/NRR-1141 L Page 131 of 285 SE No.: 89-037 Source Document: USAR CR 89-024 i Description of Change This USAR change request corrects a reference in Table 3.2-1, " Equipment Classification" under Item XXXIV, Principal Component Number 15, Spent Fuel Pool and Liner. This change request adds in the construction code RDT-F6-6T, Division of Reactor Development and Technology USAEC-Velding of Structural Components. Summary
- 1. No. There is no change to the plant. This change simply adds the proper construction code reference to Table 3.2-1, to reflect the construction code _that the Spent Fuel Pool was constructed to.
II. No. See Item I above. III. No. This is an administrative change to the USAR only. 'It does not affect :he Technical Specifications.
Attachment 3 PY-CEI/NRR-ll41 L Page 132 of 285
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SE No.: 89-038 Source Document: USAR CR 89-002
' Description of Change This USAR change request modifies USAR Section 8.3.1.1.3.3 +o eliminate the requirement to, closely monitor light load operation / service limits on the High Pressure Core Spray (E22B) Diesel Engine Turbocharger. This is due to implementation of the recommendations of NUREG CR-0660 (reference Perry Unit 1, SER Section 9.6.3.1).
Summary I. No. The USAR text described the HPCS Diesel Generator Turbocharger service limits based on the manufacturers recommendations for the originally installed standard turbocharger design. NUREG CR-0660 recommended replacement of the original HPCS diesel engine turbocharger with a heavy duty turbocharger design which was committed to in our Safety Evaluation Report (SER). GE Field. Deviation Disposition Request (FDDR) KL1-304 details this design change, installed in 1985. The engine / turbocharger manufacturer, GM/EMD, issued a bulletin which revised the light load service limits applicable to the heavy duty turbocharger: Loads Greater than 50% - 8000 hours Loads Ranging From 20% to 50% - 6000 hours Loads Less than 20% - 3000 hours This USAR change only impacts the service limits of the HPCS Diesel
' generator turbocharger, and-has no effect on plant design / configuration.
An evaluation has been performed which demonstrates that Perry's vorst case test conditions do not compromise the manufacturer's recommended service limits for the HPCS diesel engine turbocharger. Consequently, the required reliability of the HPCS diesel engine as evaluated previously is not affected by this change. Hence, the
. probability of a malfunction of equipment as evaluated previously is unchanged.
II. No. The USAR change does not introduce / create any new design configuration, nor does it reduce any equipment reliability levels. This change maintains the original design configuration and original system performance capabilities. Hence, no possibility for a malfunction of a different type is created by this change.
-III. No. The bases of Technical Specification Section 3.4-8 refers to the reliability of onsite power supplies. Item I demonstrates that this reliability is maintained, in light of this change. Hence, the margin of safety as described in the Technical Specification bases is unchanged.
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T , 4 Attachm:nt 3-PY-CEI/NRR-1141 L Page 133 of 285
. SE No.: 89-039 Source Document: SXI 0037, Rev. 0
'l Description of Change This safety evaluation examines the performance of Special Test Instruction (SXI) 0037, Standby Liquid Control (C41) System (SLC) Two Pump Relief Valve Test.
Summary I. No. The St.andby Liquid Control System is necessary to support the shutdown of the reactor independent of the control rods after achieving criticality and during power operations. The performance of this test vill occur with the reactor being in cold shutdown. The test vill use all of the C41 System components and vill not decrease the system integrity as described in the USAR. The system
. vill be tested well within the design criteria established for system operating pressures and pump parameters. To facilitate the test, the system relief valves vill be placed in a orientation different then its original design configuration. However, the over pressure protection function of the relief valves vill not be compromised. Upon completion of the test, the relief valves vill be -l' returned to their original configuration. Therefore, this test will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety that has been previously evaluated. "
'II. No. The test has been written to verify the reliability of the system i relief valves and vill not create an accident or malfunction ;
different than that described in USAR, Section 9.3.5. See Item I l above. ' III. No. This special test vill be performed with the reactor in cold ; shutdown with no control rod motion and therefore the wargin of safety.as defined by' Technical Specification 3.1.5 vill not be
.. comprised.
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I; Attachmsnt 3 PY-CEI/NRR-1141 L Page 134 of 285 SE No.: 89-040 Source Document: ISS SP-2600, SCN 00206-ISS-260 Description of Change This change elevates the addition of the ASME code case N-247 to the Installation Standard Specification (ISS) SP-2600. Summary I. No. The majority of the changes being made to ISS SP-2600 by SCN-00206-ISS-2600 are minor in nature. The addition of ASME Code Case N-247.is a change to the USAR (Table 5.2-1) and is the sole reason for this Safety Evaluation. Simply stated, this Code Case allows a manufacturer of component standard supports (such as Power Piping Company) to furnish a
" Certified-Design Report Summary" to the N Certificate llolder (such as CEI) in lieu of furnishing a " Design Report" as required by NCA-3551.1.
The certified Design Report Summary vill not change or degrade the form,-fit or function of the component standard supports being supplied to Perry. Future supports furnished by Power Piping Company will be, upon approval of SCN-00206-ISS-2600, supplied with Certified Design Report Summary sheets. The' components themselves vill continue to conform to the requirements of the existing Load Capacity Data Sheets used previously at Perry. This revision is a minor documentation change only. The use of Code Case N-247 vill not increase the probability of malfunction of component standard supports. II. No. See Item I above. Use,of Code Case N-247 vill not create the possibility of a different type of malfunction. III. No.: Safety margins-from the Technical Specifications vill not be reduced. l
f Attechmint 3 PY-CEI/NRR-1141 L Page 135 of 285 SE No.: 89-041 : Source Document: USAR CR 89-025 Description of Change This change involves reclassifying non-engineered safety features systems as engineered safety features support systems to be consistent with Section 7.3 of NUREG-0800. Summary I. No. The system classification change'does not change system operability. The safety classification of all systems and components remains unchanged. Therefore, there is no effect on the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR. II. No. There is no effect on the physical condition of systems or components. Therefore, there is no possibility of an accident of a different type than evaluated in the USAR. III. No.- The safety classification of all systems and components remains , unchanged. Therefore, there is no change to the' margin of safety. '- SE No.: 89-042 Source Document: USAR CR 89-026
. Description of Change This change incorporates reference to NRC letter, dated July 8, 1986, concerning the Transamerican Delaval Inc. (TDI) Diesel Generators, which c1cses license commitment 10(f) of USAR Appendix IB.
Summary I. No. This change simply references an NRC evaluation which closes a commitment made to the NRC. No change to the plant or procedures is being made. II. No. .This. change does not affect plant equipment, procedures, tests or experiments. There is no increase in the probability for an accident or malfunction of a different type than previously evaluated. III. No. This change does not affect the Technical Specifications or their bases.
4 Attechm:nt 3 , PY-CEI/NRR-1141 L Page 136 of 285 SE No.: 89-043
- Source Document: NR PPDS 3'251, Rev. O Description of Change This change corrects USAR Table 8.3-1 to reflect the field verified full load amperages for motor numbers OP47-B001A/B, and adjusts the Standby Diesel Generator KV loading accordingly.
Summary
-I. No. The upgrading of the OP47-B001A/B motors increased Bus loading by 1 KV. However, the 7000 KV design capacity of each standby diesel generator is more than adequate to carry this additional 1 KV automatic loading reflects in USAR Section 8.3. Therefore, an increase in the probability of an accident or malfunction important to safety previously evaluated in the USAR does not exist.
II. No.
. The motors are operating within the acceptable limits of the National Electric Code as defined per Sections 430-22 and -34.-
Therefore, the' possibility of an accident or malfunction ~of a different type than previously evaluated in the USAR is not created. III. No. The 1 KV addition vill not cause either the Division 1 or Division 2 diesel generator to exceed the 7,000 KV automatic - connected load limit defined in Technical Specification Section 3.4.8.1.1.9. Therefore, the margin of safety as defined:in the bases for Technical Specification is not reduced. l 1
-m - 7 Attcchm:nt 3 PY-CEI/NRR-1141 L Page 137 of 285 SE No.:- 89-044, 89-045, 89-046 Source Document: DCP 87-0734B, Rev. O Description of Change This design change makes pneumatic and electrical modifications to the Division I and II Diesel Generator Trip Shutdown System. Summary I. No. This change adds pressure switch ~PS-29A to the E-20 sensing line which vill maintain the engine in a shutdown condition without cranking. This circuitry change vill be capable of interrupting a ' LOCA start signal to the engine as described in USAR Section 8.3.1. The installation of this new pressure svitch vill have no affect on pressure boundary reliability 'uecause its design and qualification are consistent with the original design. The pneumatic and electrical installation standards utilized to implement this pressure switch addition are-identical to those used in the original design. This modification is a functional improvement to the DG since it prevents engine cranking following receipt of a shutdown signal. Additionally, the modification relocates the generator differential relay protection signal to the diesel generator inop/ emergency trip circuit logic. This will also maintain the engine in a shutdown condition without cranking. This circuitry change vill be capable of interrupting a LOCA start signal to the engine as described in USAR Section 8.3.1. The installation standards utilized to implement these viring/ circuitry changes are identical to those used in_the original design.
'All of the subject trips (manual, overspeed, and generator differential) are active in the event of a plant emergency. The-subject design changes do not affect the emergency response of either DG except to prevent cranking on coastdown after a trip.
llence, the subject changes do not af fect the response of either DG. to a plant emergency. Although this change adds circuits to the SSCR which is part of the Approved Fire Protection referenced in the USAR, these circuits vill be separated from the redundant trains of safe shutdown equipment and protected from the effects of fire in the same manner as previously analyzed for the fire areas involved in this change. This change vill not preclude safe shutdown during a fire nor does it alter the basis for any Appendix R deviation. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated is not increased.
r; ;~ s Attcchm:nt 3 PY-CEI/NRR-1141 L Page 138 of 285. SE No.:- 89-044,-89-045, 89-046 (continued) Summary II. No. .All of the components installed as the result of this change meet all the original equipment qualification requirements. The physical installation of the pressure switch and its associated tubing and viring as well as other shutdown / lockout viring changes is in accordance with the original installation requirements. This change creates no reduction of redundancy or component performance levels compared to the original design. Ilence the possibility for an accident or malfunction of equipment'different than previously evaluated is not created. III. No. The margin of safety as defined in the bases of Technical ; Specifiention L.stion 3.4.8 refers to the availability of the onsite power supplies. As described in Items I and II above, the operability.of the. Division 1 or 2 Diesel Generators is not changed by this design. Ilence, the margin of safety described in the Technical Specification bases is not affected. J t v
e_> - - l L Attcchment 3 : p PY-CEl/NRR-ll41 L l 1 Fage 139 of 285 1 SE No.: 89-047 l Source Document: DCN 2570 i Description of Change i This change revises Draving 808-306, Sheet 1 (USAR Figure 7.4-2, Sheet 1) ; to delete Standby Liquid Control System (SLCS) Storage Tank level inputs -! from the SLCS out-of-service annunciator. ! Summary , I . - No. The subject change to the SLCS Functional Control Diagram and T corresponding revision to USAR Figure 7.4-2, Sheet 1, are being made t for consistency with the installed SLCS field condition and other- ' Ferry design drawings. The annunciator input circuits in question I were originally added to the General Electric SLCS Function Control : A' Diagram via FDDR prior to this design being field implemented. USAR f Figure 7.4-2 was based on the CE drawing. The GE FDDR vas i subsequently canceled without appropriate drawing revisions. The annunciator circuits being deleted from the functional control diagram do not affect the SLCS design requirements. Further these. , circuits are not described in USAR Sections 7.4 and 9.3.5 or ! Chapter.15. It is concluded that the probability of occurrence or the consequences of an accident / malfunction of safety-related , equipment previously evaluated in the USAR is not increased. 1 II.-No. The scope of this evaluation is limited to the functional control _ diagram change described in 1 above. No other systems or components - are affected by the diagram revisions. A different. type of accident j or ralfunction than evaluated previously in the USAR is not created. - E
- III. No. The SLCS Storage Tank Zero Level annunciator circuits are not i described in the Technical Specifications.- Deletion of these i
circuits from the functional control diagram vill not affect SLCS '
' design and operability requirements. The margin of safety in the c bases for' plant Technical Specifications is not reduced.
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Attcchment 3 PY-CEI/NRR-1141 L Page 140 of 285 l SE No.: 89-048 i Source Document DCN 2457 Description of Change i This design change adds a note to th Emergency Closed Cooling Vater (ECCV-P42) System dravings to indicate removal of manual control unit card P42-R314B-2. This converts valve P42-F3158 from being capable of Control Room operation to local manual operation. (I&C Evaluation) Summary E L I. No. Removal of manual card P42-R314B-2 prevents Control Room operation ! of the temperature controlled motor-operated modulating valve , P42-F315B. It does not impact 1) local manual-operation of the valve, 2) other control circuits, instrumentation, ennunciators or computer points in the Division 2 train, 3) other control circuits, ; instrumentation, annunciators or computer points in the redundant Division 1 train. Since the P42-F315B valve can still be operated locally and the removal of the card does not impact the operability of other components, the probability of occurrence or the consequences of an accident or malfunction of equipment important to ! safety is not increased. ; II. No. Removal of the card only eliminates operation of the valve IP42-F315B from the Control Room and does not impact any other components. Therefore, the possibility of an accident or malfunction of a different type than previously evaluated in the USAR is not created. III. No. The ECC System operation is described in Technical Specification Section 3/4.7.1 Cooling Vater Systems and is required to ensure that , sufficient cooling capacity is available for continued operation of ' safety-related equipment during normal and accident conditions. Removal of the manual control unit card prevents Control Room valve control; however, the valve can be manually positioned to endure that sufficient cooling capacity is available during normal and accident conditions. Therefore, the margin of safety as defined in the Technical Specification basis is not reduced. k 1
Attcchment 3 e PY-CEI/NRR-1141 L ; Page 141 of 285 SE No.: 89-049 ! Source Document: DCN 2457 ! 1 Description of Change ' This design change adds a note to the Emergency Closed Cooling Vater ; (ECCV-P42) System drawings to indicate removal of manual control unit ; card P42-R314B-2. This converts valve P42-F315B from being capable of i Control Room operation to local manual operation (Mechanical Evaluation). l Summary f
- 1. No. This drawing change as-builts the conversion of IP42-F315B from [
automatic to manual operation. This change decreases the : probability of a malfunction of equipment important to safety. As described in USAR Section 9.2.2.1, the ECCW System is designed for a single active or passive failure of any component, without affecting the system's ability to perform its intended functions in Hot Standby, Normal Shutdown or Post-Accident plant conditions. This > change converts IP42-F315B from an active component to a passive component and thereby eliminates the possibility of an active l failure of this valve and its motor-operator. The Control Complex Chilled Vater (CCCV-P47) System had previously failed to start during LOOP /LOCA testing. This problem was due to the IP42-F315ABC control valves being positioned for lov cooling water temperature and the reset band of the condenser flow switch. Current system operation has establfshed proper cooling flov to the CCCV chiller by positioning the IP42-F315B control valve to establish approximately 1200 gpm to the chiller. This vill '
-satisfy the permissive circuit permitting an auto start of the CCCV i system following a LOOP or LOCA, thereby decreasing the probability of a malfunction of this safety-related equipment.
Also, this change does not increase the probability of occurrence'or the consequences of an accident of equipment important to safety for ' the reasons stated above. II. No. The possibility of excessive or insufficient cooling of the control I complex chillers by ECCV has been reviewed and determined not to exist. The purpose of the IP42-F315B valve is to restrict cooling water flow to the control complex chiller if emergency closed ' cooling vater temperature falls belov 55'F. Emergency Service Vater f (ESV) System vinter mode operation has been proven effective at 5 maintaining ECCV temperatures above 55'F vith lake temperatures of
- 55'F to 32*F. ECCW loop B Startup Test Instruction test data (3/87 e- ,
+
E t Attcchment 3 [ PY-CEI/NRR-ll41 L Page 142 of 285 . SE No. . 89-049 (Continued) [ Summary ! and 5/87, with lake temperature below and above 55'F) indicated that ! vith lake temperatures at 38'F and 57', ECCV cooling water to the ! CCCW chiller vere 93'F and 90'F respectively. Recent field data, i dated 1/18/89, demonstrated that with ESV temperature at 40'F and l the temperature control valve in manual, the ECCW cooling water to ; the Control Complex Chiller was 68'F. + 1 This demonstrates that the ESV and ECCW Systems vill provide their ; required system functions with IP42-F315B at a fixed position. ! i Therefore, possibility of an accident or malfunction of a different ;
-type than those evaluated in the USAR has not been created. ;
III. No. The CCCW system is not directly covered by Technical Specifications a but is required-to support Control Room habitability. This change vill increase the reliability of CCCW and ECCV to support this 'j function. Therefore, the margin of safety as defined in Technical t Specifications are not reduced per section 3/4.7 and 3/4.7.1 and' i 3/4.7.2. ; I [ P W r t I l t I i 5 s
l Atttchment 3 , PY-CEI/NRR-Il41 L l Page 143 of 285 ! SE No.: 89-050 t Source Document: USAR CR 89-016 i
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Description of Change , This change increases the maximum allovable normal operating dryvell temperature from 135'F to 145'F. It also revises the associated abnormal and accident dryvell temperatures. - Summary I. No. Increasing the dryvell temperature limit to 145'F vill provide the i plant operational flexibility. The design of the dryvell and i associated equipment and the accident analyses have been' reviewed to determine the effects of: ,
- a. Higher initial dryvell temperature on dryvell and containment -
responses (following small break LOCA).
- b. Peak dryvell and containment pressures. ,
- c. Higher initial dryvell temperature on structural design and l analysis,
- d. Higher initial dryvell temperature on equipment and components (environmental qualifications).
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- e. Higher initial dryvell temperature on pool swell loads.
The conclusions are that the results of the previous analyses are not affected. Halfunction of equipment is not affected because the post-accident peak temperature of 330'F is not changed.- Further, i evaluated the qualified life of the equipment when subjected .. to 145'F continuously has not been reduced. Thus, the probability-or the-consequences of an accident or malfunction of equipment is i not increased. - II. No. Raising the dryvell temperature to 145'F provides operational flexibility. Design basis accident and transient analyses, and plant system and equipment design are not effected by the change. Thus,.no new accident or malfunction is created. III. No. The Technical Specification basis for the average temperature inside $ the dryvell is to limit post accident temperature to 330'F. All design basis analyses have been reviewed and found that the limit is . not exceeded by increasing the dryvell temperature to 145'F. , 5
- Attochment 3 ,
PY-CEI/NRR-1141 L Page 144 of 285 l SE No.: 89-051 SOURCE DOCUMENT: DCN 1745 ! Description'of Change t This change updates USAR Figure 7.3-8 to be consistent vi.h the latest ! revision of P&ID D-302-832 (Hydrogen Analysis System) by adding a Master j Parts List (MPL) number reference for existing hardware. 9 i Summary ~ f I..No. Since this change concerns updating a USAR Figure to reflect an: I assigned MPL number without making hardware or system changes no ; impact to the function of safety-related equipment exists. II. No. -Assignment of an MPL number for existing hardware does not create a different type of accident than previously evaluated in the USAR. - t
'III. No. Assignment of an HPL number does not impact design or operability requirements for the Hydrogen Design Systems. Therefore, Technical !
Specification margins of safety are not involved. j SE No.: 89-052 !
, Source Document: DCN 2130 and 2232 .i USAR CR 89-031 !,
Description of Change These changes update USAR Figures 7.6-1 (Leak Detection System) and 5.4-9 (Reactor Core Isolation Cooling System) to be consistent with the latest : revision of the ascociated system P&ID. The changes update Master Parts i List (MPL) numbers and revise the location of several valves. Summary I. No. These changes are administrative only and do nut affect system design or operation, therefore the probability of an accident or malfunction of equipment is not increased. II. No. These changes are administrative only, and vill not cause an : accident different from any previously evaluated. -
.III. No. These valves are not described in Technical Specifications. These !
changes are administrative only and do not affect system operability. Therefore, no Technical Specification margin of safety has.been reduced.
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. Attechtnt 3 1 PY-CEI/NRR-1241 L l Page 145 of 285 ;
i SE No.: 89-053 through 89-065 : Source Document: See below
)
Description of Change The vendor drawings for the below listed feedvater heaters require socket l velded small bore connections. These connections are seal velded with a ! thread added to the inside for the purpose of temporary plugging during ) hydrotesting and nitrogen blanketing. Additionally, some connectors to 2 the heaters are not as shown on the detail drawings (i.e., connections in ! the heaters with or without seal velds, threaded material and plugs are of unknown composition and pressure rating, connections missing). SE No. Source Document Description 89-053 MMON 1103, Rev. 1 ' Feedvater Heater 1N21-B0002A 89-054 MMON 1116. Rev. 1 Feedvater Heater IN27-B0003B i 89-055 MMON 1115, Rev. 1 Feedvater Heater IN27-B0003A ; 89-056 MMON 1114, Rev. 1 Feedvater Heater 1N27-B0002B 89-057 MMON 1113, Rev. 1 Feedvater Heater 1N27-B0002A 89-058 MMON lill, Rev. 1 Feedvater Heater IN27-B0001A ' 89-059 MMON 1110, Rev. 2 Feedvater Heater 1N21-B0003B 89-060 MMON 1109, Rev. 2 Feedvater Heater IN21-B0003A 89-061 MMON 1108, Rev. 1 Feedvater Heater IN21-B0001C 89-062 MMON 1107, Rev. 1 Feedvater Heater IN21-B0001B , 89-063 MMON 1106, Rev. 1 Feedvater Heater 1N21-B0001A 89-064 MMON 1105, Rev. 1 Feedvater Heater IN21-B0002C 89-065 MMON 1104, Rev. 1 Feedvater Heater IN21-B0002B , Summary I. No. Condensate Sides i USAR Section 10.4.7.1 describes the condensate System, which is the e tube side of the subject feedvater heater listed. USAR ' Section 10.4.7.1.3 discusses the safety evaluation for the . Condensate System. The system is nonsafety-related. The level of radioactivity in the condensate System is lov enough that leakage vill not create hazards. To further protect the environment, all floor drains from areas where leaks could occur are piped to the liquid radvaste system for processing. The temporary "use-as-is" disposition of this NR does not increase , the probability of an accident or malfunction previously evaluated ; in the USAR. If leaks were to develop from the threaded plugs in the condensate side of the heater it would not be immediately_ felt by the reactor coolant system due to the storage capacity of the hot surge tank and the feedvater system. This vill allow for corrective actions to be taken. The threaded plugs in questions have been in service since preoperational testing began. The tube side of the subject feedvater heater was hydrotested to 1.5 times the design pressure of the heater (reference Pullman hydrotest package N21, N27). The hydrotest was in accordance with ASME Section VIII
Attcchment 3 PY-CEI/NRR-1141 L Page 146 of 285 SE No.: '89-053 through 89-065 (Continued) Summary
, Division 1 for these heeters. The threaded plugs, unidentified spare connections, and 1 inch threaded pipe mentioned in several of the NRs vill not increase the probability of an accident or malfunction previously evaluated in the USAR.
Steam Side [ USAR Section 10.2.2.1 addresses the Turbine-Generator. This section L mentions that the above heaters are designed in accordance with the L ASME code. As previously stated, the referenced plugs were not installed per ASME Section VIII. The shell-side of the following heaters were hydrotested to 1.5 times the design pressure (reference Pullman hydrotest package N25, N36). 1N21-B0003A IN27-B0002A IN21-B0003B IN27-B0002B
- IN27-B0005A I
The Extraction Steam System and the High Pressure Heater Drains and Vents System which is on the steam side of the feed heaters is e nonsafety-related. USAR Section 15.1.1 discusses loss of feedvater heating. This is related to losing the heater identified in the NRs. Loss of a feedvater heater vill cause the reactor vessel to receive cooler L feedvater and increase core power. This situation has already been analyzed in chapter 15. Section 15.1.1 addresses the consequences of an accident of malfunction of the heaters which is applicable to these NRs. If a leak developed at one of these plugs, the resultant manual isolation of the subject heater would be within the scope of this analysis. The threaded plugs were installed for hydro testing purposes. The shell side of the heaters were hydroterted to 1.5 times the system design pressure which is in accordance with A5HE Sect. VIII, Division 1 code requirements. The number 1 set of low pressure heaters operates under a vacuum pressure of 5.4 psia. A leak through one of the threaded plugs vould' affect the heaters' efficiency. Over a period time this vould affect the condenser's vacuum. USAR Section 10.4.1.4 discusses air inleakage into the condenser. The main condensers are not required to affect or support the safe shutdown of the reactor, or to support in the operation of reactor safety features.
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Attcch::nt 3 PY-CEI/NRR-1141 L Page 147 of 285 SE No.: 89-053 through 89-065 (Continued) Summary Delaying the upgrade of the deficiencies mentioned in several of the NRs, such as the referenced threaded plugs, and unidentified spare connections from the first to the second refuel outage and upgrading the 1-inch threaded pipe during the first refuel outage vill not increase the probability of an accident of malfunction important to safety previously evaluated in the USAR. The problems that could occur with the threaded connections are contained in the scope of other occurrences previously evaluated and analyzed in the USAR. II. No. The IA, B, C, and 2A, B, C heaters are located in the upper portion of the main condenser with part of the heater outside the condenser shell. All plugs identified on the corresponding NRs are located outside the condenser shell, i.e., tube side and shell side of the heater. For the 2A, B, and C heaters both the shell side and tube side of the heaters are at positive pressure. A Condensate System threaded plug blovout is not specifically discussed in the USAR. Ilovever USAR Section 10.4.7.1.3 states that leaks from the heater vill drain to floor drains and be taken to radvaste for processing to protect the environment. The USAR Section 10.4.7.1.3 addresses leaks from the , feedvater/ extraction steam heater, therefore, the possibility for an accident for malfunction of a different type than any evaluated in the USAR has not been created. The threaded plugs located on the shell side of the number 1 set of lov pressure heaters could cause condenser air inleakage. This situation has been addressed in USAR Section 10.4.1.4. Delaying the upgrading of the deficiencies mentioned in several of the NRs such as the threaded plugged connections from the first to > the second refuel outage and seal velding the 1" threaded pipe connection during the first refuel outage vill not create the possibility for an accident or malfunction of a different type than any previously evaluated in the USAR. III. No. The subject feedvater heaters are not addressed in the Technical Specifications. 9 l
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r Attech::nt 3 PY-CEI/NRR-1141 L i Page 148 of 285 < SE No.: 89-066 - l Source Document: USAR CR 89-029 i DCN 2449 i Description of Change l This change incorporates into the USAR a reformatted Figure 9.5-24, High
-Pressure Core Spray (HPCS) Diesel Generator Air Starting System.
{ Summary
-1. No. .This USAR figure change reflects the same starting air system' equipment configuration and safety classification as. defined in USAR Section 9.5.9.3 for the diesel generator. Reformatting of the :
existing Figure 9.5-24 into a new USAR figure based on
- Draving D-302-358 is an editorial / configuration improvement that vill clarify the Division 3 Diesel Generator Starting Air System.
Since there is no change to the Division 3 Diesel Generator Starting Air System as described in the text of the USAR, or as installed, the probability or consequences of an accident or malfunction is not increased. . II. No. See Item I above. ' III. No. Since this change only clarifies the HPCS diesel generator description in the USAR and that there is no physical change to the system'or its operation, the margins of safety as defined in the 'i bases of the Technical Specifications remain unaffected. P E 4
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Attach: nt 3 l PY-CEI/NRR-1141 L l Page 149 of 285 SE No. 067 I 5ource Document: DCP 88-0096, Rev. O Description of' Change l This change adds limit switches and Control Room indication for Fire Service Vater System (P54) Containment Isolation valves IP54-F726 and 1PF54-F727. i Summary-I. No. This design change adds limit switches and Control Room indication , to Manual containment Isolation Valves IP54-F726 and -F727. .It. ,l does not alter-the function or. operation ~of these valves or of the
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Fire Service Vater System. Further, this change does not impact those accidents described in_ Table 15.0-3 of the USAR. Therefore, t the probability of occurrence.or consequences of an accident or malfunction of equipment important to safety previously evaluated is . not increased. II. No. See Item I above. III. No. The manual containment isolation valves IP54-F726 and F727 are described in Table 3.6.4-1 of the Technical Specifications. However, the addition of the limit switches to the valve does not ; impact its isolation capability. Therefore, the margin of safety ; has not been reduced. I \-. i I' o I
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p i ll Atttchmnt 3 i PY-CEI/NRR-1141 L Page 150 of 285 SE No.: 89-068 L Source Document: TXI 071, Rev. 0 l Description of Change This safety evaluation analyses a Temporary Operating Instruction (TXI) concerning the use of the inlet temperature of the Residual Heat Removal (RHR-E12) Heat Exchangers (HXs) using instrument 1E12-R601 in place of the saturated water temperature inferred from the steam dome pressure to verify that less than a 50' delta-temperature (ot) exists between operating loop temperatures _and an idle recirculation loop. Installation L of temporary seal vater ::upply hoses is also evaluated. F Summary L 1. No. The criteria for starting an idle recirculation loop as stated in Section 5.3.3.6.c of the USAR, "... coolant temperature in that loop is within 50'F of the saturated water temperature corresponding to the steam dome pressure," is not being changed by TXI-0071. The purpose of the temperature restriction is to minimize thermal shock due to a pump start, to the recirculation nozzles, pumps, and piping. The use of the saturated water temperature for determining the at for the start of a loop is based on the operating plant condition where the hottest temperature in the RPV is, that of the steam dome. With the RPV depressurized and belov 212'F, as is required in condition 5, the inferred saturation temperature is not representative of the true temperature of the hottest, or any, part of the RPV. It is fixed at 212' irregardless of actual RPV temperature. The representative temperature to be used for 6t determination, with the vessel inventory undergoing continuous mixing via shutdown cooling, is that of the water in the operating shutdown cooling loop prior to entry into the RHR HXs in service. This is as close to the temperature seen by a starting recirculation loop as can be measured with installed plant equipment. Thus, the use of RHR A(B) HX'S inlet temperature in place of the inferred steam dome saturation temperature to satisfy the less than 50' at idle loop start criteria does not increase the probability or consequences of an accident or malfunction of equipment important to safety. Nor does it create the possibility of an accident or malfunction different from those already evaluated in the USAR. 3 5
1 Attcchment 3 PY-CEI/NRR-1141 L l Page 151 of 285 SE No.: 89-068 (Continued) Summary The installation of temporary seal vater supply hoses for the recirculation pump seals in Section 7.1.2 and 7.1.3 of TXI-0071-is ; accomplished using components that are not required to maintain I containment integrity and have no safety function in Condition 5. I Those components used in 7.1.3 are not safety-related and are !
' located outside of containment. Those described in 7.1.2 are l safety-related but serve a passive function in maintaining dryvell integrity which is not required in condition 5. they are located in i a suppression pool swell area and thus a hose rupture vould not I compromise any equipment important to safety. 1 The installation of temporary seal water supply hoses for the recirculation pump seals in Section 7.1.1 of TXI-0071 uses IP11-F539 l vhich is required to maintain containment integrity. Procedural .
limitations are placed on the use of this valve which prohibit its use when containment integrity is required. The valve serves a passive function and has no safety function in Condition 5 when containment integrity is not required. l Section 7.1.1 also uses IP11-F540 which is not required for containment integrity. The valve serves a passive function and has no safety function in Condition 5. Both valves are located in a suppression pool swell area and thus a hose rupture vould not compromise any equipment important to safety. , Thus the installation of temporary seal vater supply hoses by [ Section 7.1 of TXI-0071 does not increase the probability or ' consequences of an accident or malfunction of equipment important to e safety. Nor does it create the possibility of an accident or malfunction different from those already evaluated in the USAR. II. No. See Item I above. III. No.. The criteria for starting an idle recirculation loop as stated in Section 3.4.4.1 of the Bases to Technical Specifications defines a margin of 50'F between the loop temperature and "... the reactor pressure vessel coolant temperature ...". The use of RilR A(B) IlX'S inlet temperature in place of the inferred steam dome saturation temperature to satisfy the less than 50' delta-t idle loop start i criteria meets this 50' margin and thus constitutes no reduction in the margin of safety in the Bases of the Technical Specifications.
4 - Attcchment 3-
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PY-CEI/NRR-1141 L Page 152 of 285 i No. ~89-069 4 Source Documents PAP 1917, Rev. 2 ! I Description of Change l
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,The procedure is being changed because the intended use of Form No. 8340, -'
Fire Brigade Training Record, is for reference purposes only. Required l documentation of fire brigade training is in accordance with TNG-1008, i Course Completion Standards and Documentation, and through the use of the i Perry Training Section Records and Retrieval System. Vith the documentation requirements being met through TMG-1008, there is no need , for the Fire Brigade Personnel Training Record to be maintained as a " quality aJsurance record. Changes made in this procedure are.in accordance vith Section Ilof I 10 CFR 50, Appendix R. Fire Brigade Training. Summary I t See Safety Evaluation 88-470.
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m , c Attcch::nt 3 ! !-- PY-CEI/NRR-1161 L l l- Page 153 of 285
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SE No.: 89-071 l TXI-074 Source Document: Description of Change I This safety evaluation analyzes operating the Fuel Pool Cooling and , Cleanup (G41-FPCC) System under a Temporary Operating Instruction (TXI) i with the reactor cavity drained and spent fuel bundles in the upper containment pool. in the event of insufficient cooling by the normal means a contingency temporary backup system would be installed to cool the upper containment pool. Summary I. No. USAR Section 9.1.2.3.3.0, addresses storage of spent fuel in the containment pool. The storage rack design vill allow storage"of 190 bundles. The geometry of the storage rack is designed to preclude the possibility of criticality during normal and abnormal conditions. -In all cases K vill not exceed 0.95. One of analyzedboundingconditionI[sthepoolbeingdrained(Reference USAR Section 9.1.2.3.1). The vorst case of pool drainage during this operation would be if the storage pool gate is removed. Vater level above the stored fuel vould be at the step-up for the gate (Elevation 666'-4". This results in 5 feet of water being over the spent fuel. Therefore, this is within the bounds of.the safety analysis for the fuel storage racks. During the performance of this operation only 122 fuel bundles vill be-stored in the upper containment pool which would generate approximately 4.5 x 10E6 BTU /HR of heat load on the 24th day following shutdown. This heat load was based on a 1.4%' peaking factor. On day 40 following shutdown (16 days after day 24) the heat load in the upper pool vill have decreased to 75% of its day 24 value. On day 50 following shutdown heat load vill have decreased-to 66% of its day 24 value.- Calculations using the bounding case of maximum heat load show that with 300 gpm flow to the upper pool, with an Nuclear Closed Cooling (NCC) temperature to the FPCC heat exchangers of 70 degrees maximum , and with tvo heat exchangers in service, sufficient heat transfer vill take place in the upper pool. This configuration makes use of existing skimmers as a return path so that no change to the system is introduced. Return flow of 300 gpm has been observed when the upper pools vere drained previously for work. 1 i
Attcchacnt 3 PY-CEI/NRR-1141 L Page 154 of 285 SE No.: 89-071 (Continued) Summary However, in the unlikely event that insufficient cooling exists with the normal system a backup temporary return system has been developed for use to force return the vater via the skimmers to G41 and provide increased flov. This system utilizes temporary pumps that are placed in the upper pool which take suction from the pool and discharge via piping attached to the skimmers. Additionally, the pump suction location vill be above the minimum vater level considered sufficient for proper radiation shielding (23 feet). USAR Section 9.1.2.2.1 indicates that water level concerns are due to radiation shielding only. USAR Section 9.1.3.1.1 is the safety design bases for FPCC System. This section describes the FPCC as monitoring fuel pool water level and maintaining a vater level above the fuel sufficient to provide shielding for normal building occupancy. If the water level is lovered belov 688'5" (low level switch N126) an annunciator vill sound in the Control Room. any further lowering of water level vill not be monitored by G41. Level must be determined to be at least at its minimum depth once per 7 days, per Technical Specification 3/4.9.9. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. Any malfunction ir.volving draining of the pool vould be bounded by the case described in Item I above. Any failure of installed permanent equipment vill result in the same scenario as is presently evaluated in the USAR. In any case involving loss of cooling to the upper pool due to the temporary equipment, cooling may be re-initiated by flooding the upper dryer pool to have water run over the vall separating the dryer pool and reactor cavity and removing - the installed blanks over the covered skimmers. All accidents involving a dropped bundle and the subsequent release of radioactive noble gasses do not take credit for pool volume. They assume an instantaneous release of the gasses to the containment atmosphere. No pool volume is specified. Therefore, performance of this activity will not create an accident of a different type than previously evaluated in the USAR. III. No. The bases for the Technical Specifications water level in the upper containment pool ensures that sufficient vater depth is available to remove-99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assemble. The install, tion of the , temporary pump vill be such that sufficient depth vill be maintained and consequently there vill be no reduction in the margin to the basis for T. S. 3/4.9.9. Furthermore during this period of storage there vill be no movement of irradiated fuel bundles so that there is no likelihood of rupturing an irradiated fuel bundle by dropping it.
U f Attcch::nt 3 PY-CEI/NRR-1141 L L Page 155 of 285 o SE No.: 89-072 Source Document DCN 2594 Description of Change This change adds a note to the Standby Liquid Control (C41) System P&ID (USAR Figure 9.3-19) clarifying when the $5-gallon drum at elevation 620'-6" is to be removed. Summary. I. No. The Standby Liquid Control System safety function is to provide reactivity control in the unlikely event of not enough control rods being available for insertion into the reactor to accomplish shutdovn and cooldown in the normal manner. The SS-gallon removable drum is used only for the draining of the plunger pot drain and the , test tank during testing and/or maintenance activities. Removal of the drum is allowed for in the original system design as shown in Figure 9.3-19 of the USAR. Specifying removal during operating modes 1, 2 and 3 vill insure that there is no possibility of the tank becoming a missile during a reactor blovdown and subsequent suppression pool swell. Therefore, this change reduces the probability of occurrence and consequences of an accident or malfunction that may affect equipment important to safety. II. No. Due to the drum being an originally removable item, specification of the operation modes during which the drum shall be removed vill not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR. III. No. The Technical Specifications Bases Section 3/4.1.5 makes no implied or direct mention of the drum. Therefore, the margin of safety as defined in the bases of the Technical Specifications are not affected. r
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Page 156 of 285 SE No.:' 89-073 . Source Document: NR PPDS 3622, Rev. 1 ( Description of Change f This safety evaluation analyzes effects on system operation (and provides justification for continued operation) of the Puel Pool Cooling and i Cleanup (G41) System with snubber IG41-H1009 incapable of being ; activated. Field Change Request (FCR) 7542 requires all snubbers ' installed above the 654' elevation in the dryvell to be stroked to verify free movement. i summary I
- 1. No. The piping in question is one-inch in diameter. Section 3.6.2.1.6 -
of the USAR does not postulate breaks for piping with a nominal ! diameter less than or equal to one inch. However, accident analysis studies the possibility of breaking a one-inch line at the Reactor Vessel Flange with a subsequent loss of RPV coolant inventory. If ! the line vere to be completely severed, the flow from the line vould- I be 52 gpm. Make-up capacity available from condensate Transfer and 1 Storage system or Emergency Service Water system would exceed the flow loss from this event. Upper pool level vill be maintained at ' normal levels, allowing fuel assemblies to be maintained in safe wondition. . Thus, the loss of RPV coolant inventory is not a valid ; concern. Therefore the probability or occurrence of an accident or ' malfunction of equipment Important to safety previously evaluated is not increased. II. No. See Item I above. III. No. The margin of safety as defined in this bases for Technical Specifications 3.4.9.8 and 3.4.9.9 is not reduced. The water level of the upper pools vill not be affected. RSE No.: 89-074 Source Document: NR PPDS 3631, Rev. 1 i Description of Change I This safety evaluation analyzes effects on system operation (and provides justification for continued operation) of the Fuel Pool Cooling and Cleanup (G41) System with snubber IG41-H1026 incapable of being . activated. , Summary See Safety Evaluation 89-073. 4 r ~ w -
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Attcche:nt 3 i PY-CEI/NRR-1141 L i [ Page 157 of 285 ) SE No.: 89-075 'i Source Document: MFI 1-89-082 i Description of change This change installs orifice plates at the Containment. Vessel Cooling _(M11) System Air Handling Units to allow the fans to run during . the Containment Integrated Leak Rate Test (ILRT). These orifice plates- l vill be removed prior to plant startup following the first Refuel Outage. Summary
- 1. No. The Containment Vessel Cooling System (M11) is a nonsafety system and can not increase the probability of occurrence or the ;
consequences of an accident. II. No. The addition of the orifice plate does not' create the possibility of an accident or malfunction since it vill only reduce airflow through ; a nonsafety HVAC system. III. No. Technical Specification 3/4.6.1.7 on Containment Average Air .;
; Temperature does not apply to Operational Conditions 4 and 5. 3 Therefore, the margin of safety is not affected by operation of M11 in modes 4 and 5. ;
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Attcchment 3 PY-CEI/NRR-1141 L Page 158 of 285
.SE No.: 89-076 ,
Soutie Document: DCP 87-0734, Rev. 0 Description of Change This design change modifies the Division 1 Diesel Generator (DG) pneumatic control / air supply systems for overspeed trip shutdown, emergency stop shutdown, and normal engine shutdown. . Summary ! I. No. This design change combines the emergency stop signal with the overspeed trip signal. This provides a rapid response overspeed trip shutdown to protect the engine from possible damage in the event of a governor failure at full fuel rack position. This combined trip signal remains active in the event of a plant emergency. The change does not affect the emergency responce of the DG. All components and air piping associated with this change are equivalent to that of the original design. Overall, this design change does not alter the function or operation of the Division I Diesel Generator as described in the USAR. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated is not increased. II. No. All of the components installed as the result of this change meet all the original equipment qualification requirements as described in Item I above. The physical installation of the new components and their associated tubing modifications are in accordance with the original installation requirements. This change creates no reductions of redundancy or component performance levels when compared to the original design. Hence, the possibility for an accident or malfunction of equipment of a different type than , previously evaluated is not created. III. No. The margin of safety as defined in the bases of Technical Specification Section 3.4.8 refers to the availability of the onsite power supplies. As described in Items I and II above, the operability of the Division I Diesel Generator is not compromised by this design change. Hence, the margin of safety described in the Technical Specification bases is not affected. P 1 i 4
pr Attcchment 3 PY-CEI/NRR-ll41 L Page 159 of 285 i SE No.: 89-077 [ Source Document: DCP 88-0293, Rev. O Description of Change This design change relocates the four existing Heater Bay Ventilation t. L System freezestats across the four banks of coils, adds a freezestat for the fifth coil, provides a new separate trip logic, and adds bypass switches for each freezestat for maintenance purposes. F Summary f I. No. The Heater Bay Ventilation System Heating coil freezestats perform a L nonsafety trip and alarm function. Hodifications to this function cannot increase the probability or consequences of an accident. This modification enhances the effectiveness of the system and therefore does not increase the probability or consequences of a malfunction of equipment important to safety. II. Ne. Modification to the heater coil freezestat logic cannot create a new disturbance that would affect the reactor core or coolant pressure
? boundary. Therefore, the possibility for a new accident or l malfunction does not exist.
III. No. The Heater Bay Ventilation System freezestat logic is not described in the Technical Specifications; therefore, the basis is not impacted. l l l~ l l l i
Attochment 3
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l Page 160 of 285 SE No.: 89-078 Source Document: NR PPDS 3650, Rev. O Description of Change This safety evaluation analyzes the affect on operation of the Control Complex Vater Chiller (0P47-B001B) with a pin hole leak in a veld in the Hot Gas Bypass Line. i Summary I. No. This NR permits interim operation of the P47 "B" chiller until a , permanent repair (revork) can be performed. The chiller vill function / operate as described with a temporary repair provided the refrigerant levels are maintained and verified every four hours. In addition, the environmental review has shown no adverse affects to the plant and the environment. The affect on the M21 system charcoal filters are transitory with no long term impact. These factors coupled with the precautions on the torporary repair vill render the system operable. Based on the fact that system function / operation has not changed /affected, the probability of occurrence or the consequences of an accidetit or malfunction of equipment is not increased. II. No. The interim operation of the P47 system vith the temporary repair vill permit the system to function / operate as designed. Thus, malfunctions of a different type are not created. III. No. Due to the small size of the refrigerant leak the chiller can still a perform its intended function. Also, there is ample time to replenish the refrigerant without causing adverse conditions in the ' ESF HVAC systems served by the P47 chilled water system. Therefore, the margin of safety of the HVAC systems which are addressed in the Technical Specifications are not reduced. l l
( l Attcchment 3 t PY-CEI/NRR-1141 L ' Page 161 of 285 ; SE No.: 89 079-Source Document: DCP 88-068, Rev. 1 , Description of Change : This design change installs interlocks to automatically open the Emergency Service Vater (ESV-P45) inlet and outlet valves (IP45-F014A/B and IP45-F068A/B) to the Residual Heat Removal (RHR-E12) System Heat Exchangers when the associated Emergency Service Vater Pump IP45-C001A/B starts, i Summary u I. No. During normal plant operating valve line-ups, the RIIR heat exchanger inlet / outlet valves are closed and do not " auto" open on start-up of the applicable ESV pumps. With the start-up of tl.ese pumps in parallel with the close position of the specific RHR heat exchange inlet / outlet valves, flow is restricted and results in a lover flow rate while increasing the head pressure of the ESV pumps. This creates an overload condition for the pump (s) which in turn causes higher current and temperatures in addition to the lov flov vhich is approaching shut-off head pressure / flow. This DCP revision provides for the " auto" open of the respective outlet valves associated with each of the RHR Heat Exchangers. In this manner higher flow rates vill be experienced along with normal temperatures and running amps for the respective pump. This vill eliminate unnecessary or spurious trips as has been previously identified. Addition of this " auto" open logic vill not degrade the operation or affect any functions previously designated; therefore, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. Ho. This design modification prevents unwanted tripping of the ESV pumps. No new components are added. All viring is Class lE and meets the separation criteria. These modifications have not affected the other functions of the valves or the pumps. Therefore no new accident or malfunction of a different type than previously evaluated / discussed in the USAR is created. III. No. This change does not affeet Technical Specifications 3/4.7.1. Therefore the margin of safety will not be reduced. .
r m- , t l Attcchment 3- I PY-CEI/NRR-1141'L .! p Page 162 of 285 SE No.: 89-080
., _; Source Document: DCN 2579 1
Description of Change j This change corrects notes on environmental condition drawings (USAR l Figures 3.11-13, 3.11-20 through 3.11-27, 3.11-31 and 3.11-35). ! e : p Summary i j I. No. Changes to the environmental drawings are only editorial, correcting. .; design information which was transposed in error when the drawings . F vere created. Therefore, the consequences of an accident or
! malfunction previously evaluates in the USAR is not increased. j i II. No.. Corrections to the drawings do not change any design information. --
- j. Thus, it does not create any new accidents or malfunctions. t i III. No. The design numbers which are the bases to the. Technical -!
Specifications are not affected by this change. , t
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i~ l Attcch::nt 3 PY-CEI/NRR-1141 L Page 163 of 285 SE No.: 89-081 Source Document: DCP 89-0044, Rev. O Description of Change Modifications to the Unit 1 Cooling Tower to enhance performance. These include replacement of most of the nozzle assemblies, additional holes drilled in existing vater distribution piping to accommodate additional nozzle assemblies, polyvinylchloride (PVC) extensions of existing
, asbestos-cement (AC) pipes, addition of PVC lateral pipes at Flume 2 and 3, addition of stainless steel vater flow diverters, addition of fiberglas panels and supports where required, and the addition of PVC fill on top of existing AC fill in the center of the tover.
(Mechanical Evaluation). Summary I. No. USAR Section 10.4.5.2 addresses the Circulating Vater (N71) System. System analysis has shown that failure of the Circulating Vater System vill not compromise any safety-related systems or prevent safe shutdown. This design change upgrades the water and air distribution system within Unit 1 cooling tover, which is a part of the circulating water system. The modifications to the cooling tower vill not change its original function. The modifications are intended to increase the cooling tover's capability by approximately 15% to - return it to its original design. The modifications vill not affect the postulated circulating water flooding events as discussed in USAR Sections 10.4.5.3.1, 10.4.5.3.2 and 2.4.13.5. Further the changes-to Unit 1 cooling tower vill not affect any of the designed preventions of any injection of radioactive material into the circulating water system as described in USAR Section 10.4.5.3. The modifications are contained within Unit I cooling tower. There vill be no additional height or vidth added to the tover. Therefore, there is no increase of the potential for debris damaging any plant structure should the tower collapse as described in USAR Section 10.4.5.3. l The cooling tover modifications vill not affect any of the design basis accidents described in USAR Chapter 15. Cooling tower fill vill be added per this change. The fill material is PVC. This I material is combustible and is in disagreement with USAR Section 10.4.5.3 vhich states "the cooling tower is made entirely of noncombustible material". There is only a small amount of PVC fill l material being added relative to the total amount of fill. The l added PVC fill vill pose no significant fire hazard as determined by the Fire Protection Safety Evaluation.
g -- - Attechment 3 ! PY-CEI/NRR-1141 L Page 164 of 285 , SE No.: 89-081 (Continued) Summary Therefore, there is no increase in the probability of occurrence or l consequence of an accident or malfunction of equipment important to j i safety previously evaluated in the USAR. d l II. No. The original function of Unit 1 Cooling Tover v111'not-be changed as *
, a result of the modifications added. Although the modification does ,
introduce combustible materials to the cooling tower, the ! protection.of these combustibles are within the scope of the Fire 'l
; Protection Program. Therefore, no accident or malfunction of a :
different type than previously evaluated in the USAR is created. ' III. No. .The Circulating Water System is not addressed in the Technical i- Specifications. The modifications vill not change the original . l function of the Unit 1-Cooling Tower. There vill be no decrease in - s the margin of safety as defined in the basis of the Technical ; L Specifications. ,
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. Attcchment 3 L" PY-CEI/NRR-ll41 L t Page 165 of 285 j SE No.: 89-082 Source Document: DCP 89-0044, Rev. O Description of Change Modifications to the Unit 1 Cooling Tower to enhance performance. See Safety Evaluation 89-081 for a general description of this change. This safety evaluation analyzes the affects of adding the new polyvinylchloride (PVC) plastic type fill material which results in a i- combustible type construction for the Unit 1 Cooling Tower which was previously evaluated in the USAR as noncombustible construction. (Fire Protection Evaluation)
Summary
- 1. No. USAR Section 9A.5 states that the cooling towers are of
- noncombustible construction. This resulted in no fire hazard exposure to safety-related buildings from these structures. This 7 change installs additional fill material to the Unit 1 Cooling :
Tower. This new fill material is a PVC plastic which is a combustible material and changes the fire hazard of the cooling tover. The new fill vill be located inside the concrete cooling tower valls above the existing concrete fill and vill be installed I foot deep over a 39,000 square foot area with a second one foot layer installed over 1650 square feet. Although the total amount of material introduced is high (approximately 120,000 lbs) it represents a small percentage of the overall material in the cooling tover. The new fill material was tested in accordance with the ASTM E-84-84A methods and has a flame spread of 5 which is lover than the maximum flame spread of 25 allovable in safety-related , buildings. The nearest major building is the Circulating Vater Pump flouse located over 100 feet away from the cooling tover. This is a noncombustible building'vith a lov hazard occupancy and contains no safety-related equipment. The fuel oil storage tank is over 200 feet from the tower. The cooling tower is over 500 feet from any plant building containing safety-related equipment. The buildings and storage between vould not result in a spread of fire from the cooling tower to these buildings. The cooling tower basin does not serve as a water supply to any safety-related systems or serve as the ultimate heat sink. The Emergency Service Water System supply comes from Lake Erie which serves as the ultimate heat sink. The water supply for the fire pumps is also from the lake. The cooling tower basin is not connected to or required for any fire protection systems. Vith the lov hazard introduced by the new fill, separation from safety-related equipment, reliance on another source for fire suppression water and the ultimate heat sink, the introduction of a
r-Attcchment 3 PY-CEI/NRR-ll41 L g Page 166 of 285' SE No.: 89-082f(Continued) Summary L combustible construction material into the Unit 1 Cooling Tower vill not increase the probability of a fire in safety-related buildings or increase the damages or effects of a fire to equipment in these buildings.
; . 11.=No. Although this change does introduce additional combustibles the only , accident-vould be a fire and protection of these combustibles is within the scope of the Fire Protection Program. Therefore, the possibility of an accident or malfunction of a different type than evaluated in the USAR is not created.
III. No. Only administrative aspects of the Fire Protection Program and Alternate Shutdown Capability are defined 1 1n the Technical. Specifications. This change has no impact on the administrative aspects of the Fire Protection Program and does not adversely affect Alternate Shutdown Capability. Therefore, the margin of safety as defined in the basis for the portions of the Technical Specification applicable to fire protection vill not be affected. i. L L 1 o ,
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'Attcch;snt 3 PY-CEI/NRR-1141 L I Page 167 of 285 w >
SE No.: 89-083 Source Document: PAP-1923, Rev. 1, TCN 1 C o' Description of Change Temporary Change Notice (TCN) to PAP-1923, Action on Inoperable Fire Protection Systems, changes responsibilities and stated compensatory actions from those stated in Revision 0. 9 Summary
.s I. No. This procedure change is administrative in nature. As such it does t.
.not impact the accident analysis described in the USAR. .Therefore, ;
the probability of occurrence or consequence of an accident or
- alfunction of equipment- important to safety previously evaluated is not increased.
II. No. This procedure change is administrative it. r6ture. It does not affect the USAR Fire Hazards Analysis (Section 9.5.1 and-Appendix 9A). Therefore, the possibility for an accident or
-malfunction of a different type is not created.
III..No. The organizational and responsibility changes addressed by this l procedure change vill not impact the Fire Protection Program nor the alternate shutdown capability. Therefore, it vill not reduce any margin of safety defined in the Technical Specifications. b t l? 01-l
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; Page 168 of 285 SE No.: 89-084 Source Document: MFI 1-89-059 Description of Change Installation of a locking mechanism on the oartice Air System Containment Isolation Valve IP51-F150 to maintain the valve 9 pen during the 7
Division I electrical outage. Summary I. No. Containment Isolation Valve IP51-F150 is not required to be operable l in Mode 5 for when not handling irradiated fuel in containment. The valve fails closed on loss of power which ensures containment isolation. The plant vill remain in mode 5 (not handling irradiated 7 fuel in containment) during this electrical outage, Hence, the probability of the occurrence or consequence of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. Containment isolation vill not be needed during the performance of the Division 1 outage. The locking mechanism vill only be utilized during this time. Therefore no new accidents or malfunction i possibilities d e created. III. No. Technical Specifications do not. require IP51-F150 to be operable in mode 5 nor when not handling irradiated fuel in containment. Therefore, no Technical Specification margin to safety is reduced. l l E , i L
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4 !L Attcchzent-3 PY-CEI/JRR-1141 L Page 169 of 285 SE No.: 89-085 ' Source Document: HFI 1-89-097 Description of Change 1 Temporary air supply is required during the refueling outage while MOVATS testing of Service Air (PSI) System Dryvell Isolation Valve IPSI-F652 and-the Local Leak Rate Test (LLRT) of Containment Penetration P-308 are in
- progress.
i Summary-I. No. P51 supply vill be isolated to containment for a short duration during the refueling outage. Temporary air supply is necessary inside the containment to perform H0 VATS testing vork on IP51-F562 and the LLRT of Containment Penetration P-308. In addition, P51 air supply is currently being used to pressurize the inflatable seals on the main steam line plugs during the refueling outage and therefore interrupted service of P51 is unacceptable. Therefore, installation of temporary air jumpers inside containment are needed during that , time. The cross-tie occurs in the distribution piping of the Service Air and Instrument Air (P52) Systems inside containment. Upon completion of work (estimated to be less than one veek) P51/P52 vill be restored to.their original configurations. The analysis in USAR Section 15.2.10 is applicable to loss of instrument air during plant " normal" power operation. 'Therefore, the probability of occurrence or the consequences of an accident or
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malfunction of equipment important to safety previously evaluated in ; l the USAR is not increased. 1 II. No. The possibility for an accident or malfunction of a different type than previously evaluated in the USAR is not created. Air pressure is the same since both systems share common air compressors. Ilovever, PS2 air being temporarily supplied to the P51 distribution piping is of a higher quality than that of PSI, since it'is routed through the prefilters, air dryers and after filters. Therefore, , degrading or detrimental effects on the P51 System are not probable. P51 can be at a slightly higher pressure than Instrument Air (PS2) since it is not routed through the filters and dryers i.e., no pressure drops vould occur as is possible during.the~ time that both systems are in service and cross-tie valves are open. To avoid the potential of P51 air contaminating PS2 distribution piping, check valves vill be installed in jumper lines to prevent this possibility. III. No. Temporary air jumpers do not af fect any Technical Specifications. Therefore, no margin of safety as defined in the bases for any Technical Specification is affected.
pr , L Attechm:nt 3 t PY-CEI/NRR-1141 L Page 170 of 285 iSE No.: 89-086
-Source Document: HFI 1-89-155 Description of change Temporary air. supply from the Instrument Air (PS2) System to the Service Air:(PSI) System is needed to supply various P51 loads during the i Refueling Outage when P51 is taken out of service to perform installation of main header isolation valves, n -Summary I. No. Temporary air supplies are required during the refueling outage to allow installation of main header isolation valves. Installation requires shutdown of nearly the entire P51 System. Jumpers vill provide the necessary P51 services until P51 is put back in operation. P51 air supply is currently being used in the containment to pressurize the inflatable seals on the main steam line plugs therefore interrupted service of P51 is unacceptable.
The containment cross-tie connections occur in the distribution piping of the P51/P52 Systems both inside and outside of containment. Upon completion of work P51/P52 vill be restored to their original configurations. The analysis in USAR Section 15.2.10 is applicable to loss of instrument air during plant " normal" nover operation. The plant is l in mode 5 and vill remain in this condition during the performance ! of the main header isolation. Therefore, the probability of ' occurrence or the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the USAR is ! not increased-due to the shutdown status of the plant, j 7 II. No. The possibility for an accident or malfunction of a different type than previously evaluated in the USAR is not created. Air pressure is the'same since both systems share common air compressors. However, PS2 air being temporarily supplied to the P51 distribution piping is of a higher quality than that of P51, since it is routed-througn the prefilters, air dryers and after filters. Therefore, i degrading or detrimental effects on the P51 System are not probable.
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P51 can be at a slightly higher pressure than PS2 since it is not l routed through the filters and dryers. Therefore, the tendency for P51 air to back flow into P52 piping is possible during the time that both systems are in service and cross-tie valves are open. To J avoid the potential of PSI air contaminating P52 distribution piping, check valves vill be installed in jumper lines to prevent this possibility. III. No. Temporary air jumpers do not affect any Technical Specifications. Therefore, no margin of safety as defined in the bases for any Technical Specification is affected.
Attcchm:nt 3 + PY-CEI/NRR-1141 L Page 171 of 285 SE No.: 89-087 -
, Source Document: HFI 1-89-091 ;
Description ~of Change This safety evaluation is performed for the installation of temporary I operating devices required to permit operation of the Containment and Dryvell Purge Ventilation.(M14) System while rebuilding the actuators for dampers 1H14-F040, 1H14-F090, 1H14-F190, 1H14-F195, 1H14-F200, and 1H14-F205. The operating devices vill permit t.he blocking of the associated damper in either the open or close position. System operation vill be administratively controlled by a Temporary Operating Instruction (TXI). Summary I. No. With the plant shutdown in Operational Condition 5 or "at all times" s and with no irradiated fuel movement in the containment building, i the requirement for containment integrity.is no longer necessary. - The M14 dampers are evaluated for use as containment integrity isolation barriers. Since containment integrity is not required, the M14 dampers are not required to be operable. This would permit performing maintenance on the dampers. However, temporary operating devices vill'be installed which are administrative 1y controlled to ensure the dampers could be opened / closed. Therefore there is no : increase in the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. II. No. See Item I above. III. No. Temporary conditions vill only be in place during conditiuns 5 and "at all times" when containment integrity-is not required. Therefore the margin of safety as required by the Technical Specifications is maintained. (Technical Specifications 3.6.1.1, 3.6.1.8, 3.6.4, and 3.3.2). 1
Attech::nt 3 PY-CEI/NRR-1141 L Page 172 of 285 q .SE'No.: 89-088 Source Documents LL&JED 1-89-092 Description of Change , This safety evaluation is performed for the installation of temporary jumpers required to permit operation of the Containment and Dryvell Purge Ventilation (M14) System while rebuilding the actuators for dampers 1H14-F040, 1H14-F090, 1H14-F190, 1H14-F195, 1H14-F200, and IM14-F205. The jumpers vill defeat the interlocks between the fans and the dampers due to the removal of the limit switches which provide these interlocks during the rebuild. System operation vill be administrative 1y controlled by a Temporary Operating Instruction (TXI). Summary See Safety Evaluation 89-087.
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, Page 173 of 285-SE No.: 89-089 i Source Document: DCP 86-0629H, Rev. O ;
Description of-Change Modify the Containment Vessel Chilled Vater (2P50) System supply line to-regulate flow to'the Measuring and Test Equipment (M&TE) Hot , Shop Miscellaneous Ventilation (M46) System Room Cooling
- Coil (M46-B0020) by adding a temperature controller and 3-vay bypass valve.
Summary l
- 1. _ No . The Unit 2 In Service Inspection Room has been converted to a ,
Measuring and Test Equipment (M&TE) Hot Shop for use in Unit 1 plant ! operation. This Design Change vill provide the temperature controls L for the M&TE Hot Shop. The change consists of the installation of a temperature thermocouple, temperature controller, and 3 interconnecting viring inside the M&TE Hot Shop. This DCP also provides the installation of a 3-vay mixing valve, 1-inch bypass i line and orifice plate to regulate the chilled vater (0P50) flow to the Maintenance and Test Equipment Hot Shop Room Coil (2H46-B020) located in the Intermediate Building 654 foot evaluation. 1 The OP50 system affected by this DCP is nonsafety-related. The above modification has no effect on the safety-related containment Isolation Valves of the P50' system. The design modifications do not increase the consequences or occurrences of an accident or 1 malfunction of safety-related equipment nor do they create-for the L possibility of a different type than previously evaluated in the l USAR. II. No. See Item I above. III. No. The portion of the OP50 system which is bound by the Technical Specifications is the Containment Isolation Valves. The -s l modification to the OP50 system does not affect the Containment l Isolation Valves in any way_nor does it change the operation of the plant as described in the Technical Specifications. Therefore the margin of safety as defined in the bases for the Technical-Specifications has not been reduced.
E Attachmsnt 3 PY-CEI/NRR-1141 L Page 174 of 285 SE No.:' 89-091
' Source Document: DCP 87-07910, Rev. O Description of Change This design change reorients the relief valves 1R44-F0508A/B and F0518A/B on the Division 1 and 2 Diesel Generator starting air receiver tanks to a vertical position. Testing and evaluation of the currently installed relief-valve design has consistently shown the valves to be less susceptible to actuation as the result of impact or other high acceleration when oriented such that the stem axis is vertical versus horizontal.
Summary I. No. This change orients the relief valves to a position which has been demonstrated.by. testing to exhibit greater resistance to incidental actuation initiated by impact or vibration. Although these' valves have been demonstrated to maintain a closed position under vorst case accident / operating loadings in their current horizontal orientation, this change vill enhance their existing level of reliability by minimizing the potential for incidental actuations. Further, the change is demonstrated by calculation to maintain the previously evaluated factors of safety for all operating conditions. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. II. No. As described in Item I above, the possibility for an accident or malfunction of a different type than previously evaluated is not created. III. No. The margin of safety as defined in the bases of Technical Specification Section 3/4-8 refers to the reliability of the onsite power supply. As described in Item I above, the reliability of the Diesel Generators remains unaffected by this change. Therefore, the margin of safety as described in the bases of the Technical Specifications is not reduced.
Attech2:nt 3 PY-CEI/NRR-1141 L Page 175 of 285 SE No.: 89-092 Source Document: -TXI-0074 Description of Change
~This safety evaluation reviews allowing operation of the Fuel Pool-Cooling and Cleanup (G41-FPCC) System in a less restrictive lineup than was previously reviewed by Safety Evaluation 89-071.
Additionally, if needed, the operation of the temporary pumps
; (per TXI-0074) may take place with C41 operating in a configuration other than specified in Safety Evaluation 89-071. The flows and temperature restrictions of the two G41 pumps, two heat exchangers, and 70 degrees maximum inlet Nuclear Closed Cooling System temperature to the heat exchangers outlined in Safety Evaluation 89-071 vere based on calculated values of heat load.- These values were extremely conservative. Actual system operating data over a 27 hour period shows that the heat load and ,
consequently the cooling needed is much less than was calculated. It is therefore not necessary to operate in the manner specified in Safety Evaluation 89-071. This safety. evaluation supplements Safety Evaluation
- i. 89-071 in light of actual operating data (versus calculated data) and I
allows operation of the G41 System in any configuration necessary within the bounds of approved operating instructions to maintain the upper pool temperature less than the 127 degree maximum temperature specified in the USAR. Summary
- 1. No. See Safety Evaluation 89-071.
II. No. See Safety Evaluation 89-071. III. No. Technical Specification margins of 23 feet of water over irradiated fuel stored in the upper pool are not affected by operation of the G41 System in a 2 pump, 2 heat exchanger mode; a 1 pump, I heat exchanger mode, or a no pumps mode. l
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'PY-CEI/NRR-1141 L b Page 176 of 285 SE No.: 89-093 Source Document: OH16F Physical Security Plan, Rev. 13 7 . . -Description of Change Revision 13 of OH16F physical Security Plan,.Rev. 13 has been evaluated to ensure that the effectiveness of the Perry Nuclear Power' Plant Security Plan has not been reduced and to ensure that these changes meet' the requirements of the 10 CFR, Part 73, Physical Protection of Plants .f and Haterials. . Site Protection must be contacted for further details because'information is considered " safeguarded."
Summary I. No. 0H16F describes the comprehensive Physical Security Program and therefore, does not affect the occurrence or consequences of an accident or malfunction of equipment. II. No. OH16F does not direct the operation of plant systems or equipment- . and, therefore, does not create the possibility for an accident or ! malfunction. I III. No. OH16F does not reduce-the margin of safety as defined in the basis ! for any Technical Specifications. , i i I
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Attach:2nt 3-PY-CEI/NRR-1141 L Page 177 of 285 SE No.: 89-094 Source Document: DCN 2622, Rev. 0 O . Description of Change.
'This change corrects relief valve set pressures listed on drawing D-302-141 (USAR_ Figure 10.1-10).
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Summary I. No. This DCN vill not change the function of the Steam Seal-System as ,t described in Section 10.4.3. The DCN simply corrects the relief' ' valve set points on drawing D-302-141 to agree with the vendor drawing and the field conditions. This DCN vill not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. II. No. See Item I'above. III. No. This DCN. vill not' affect the operability or availability of any system described in the Technical Specifications. Therefore, no margin of safety as defined in the bases for any Technical Specification is reduced.
k Attcchtint 3 PY-CEI/NRR-1141 L [ Page 178 of 285 SE No.: 89-096 Source Document: DCP 87-0301, Rev. O Description of Change This design change adds a 120 VAC power relay to the circuitry of 1G33-F028 in series with the valve limit switch in order to disable the
. Reactor Vater Cleanup (RWCU) System Blowdown Line Hi/Lo Pressure Alarm (1H13-P680-1A-1C). This is accomplished by opening a contact in series with the associated annunciator logic when valve 1G33-F028 is l- closed.
Summary l
.I. No. This change does not alter the operation / function of 1G33-F028 or the operation of the RUCU System. The added 1G33-F028 circuitry is electrically isolated with a safety class 120 VAC power relay from the corresponding annunciator circuitry which only affects the 1H13-P680-1A-AG annunciator logic. .
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.The RVCU Blovdown Line Hi/Lo Pressure Alarm is used only for operator information.- It is not part of any accident or malfunction of equipment evaluation in the USAR. Therefore, the proba'oility of .j the occurrence or consequences of an accident or malfunction of !
equipment important to safety previously evaluated in the USAR is . not increased. l II.-No. See Item I above. i III.-No. The RUCU Blowdown Line Hi/Lo Pressure Alarm logic does not affect I the bases for any Technical Specification, j
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PY-CEI/NRR-1141 L Page 179 of 285 SE No.: 89-097 Source Document: PAP-1911, Pire Emergency, Rev. 2
. Description of Change This evaluation examines changes to PAP-1911, Fire Emergency. Changes were made under the responsibility section of the procedure for clarification purposes and to address the recent changes that have occurred in the PNPP Fire Protection Organization.
Changes made to this procedure vere done for clarification purposes. This. change is an improvement to the current method of operation and vill
-develop consistency in Fire Brigade response and the location of Fire Brigaee equipment, chich is addressed in PAP-1919, Fire Brigade Stations and Equipment. j Summary I. No. All changes made to this procedure were evaluated with respect to the USAR. All changes made vere found to be consistent with-the l fire protection requirements of the USAR. Therefore, the l probability of an occurrence or the consequences of an accident !
previously evaluated in the USAR was not increased. II. No. All changes made to this procedure have been evaluated and found'to be consistent with the fire protection requirements of the USAR.
.Therefore, the possibility of an accident different from any ,
previously evaluated in the USAR has not been created. i i i' III. No. The Fire Protection Program is referenced under administrative controls section 6.5.1.6N, 6.5.2.8E, 6.8.1H, of the Technical Specifications. All changes made in this procedure were evaluated ; and'found to be consistent with these Technical Specifications _j Section and the USAR. Therefore, no margin of safety has been reduced.
)
i l
n ~ E l Attcchm:nt 3 1 PY-CEI/NRR-ll41 L I Page 180 of 285 SE No.: 89-098 Source Document: SOI G36, Rev. 4; TCN-006 L Description of Change Temporary change Notice (TCN) to the system operating instruction of the Reactor Vater Cleanup (RVCU) System requires the RUCU Filter / Demineralizer precoat cycle to be performed in a manner which exceeds one , hour. .USAR Section 5.4.8.2 needs to be updated for it currently states , i, that the backvash and precoat cycle takes less than one hour. Summary I. No. The extended timing of the precoat cycle by this TCN-006 is intended to improve the quality of the water in the Reactor Pressure Vessel. It vill not alter the function / operation of the RUCU System. Since water quality is improved there vill be no change to the probability of occurrence or the consequences of an' accident previously analyzed in the USAR. II. No. See Item I above III. No. The longer precoat time implemented by this TCN changes no margin of. safety defined in Technical Specifications 3/4.4.4, 3/4.4.5, since no chemistry limits were changed.
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Attechmint 3 PY-CEI/NRR-1141 L Page 181 of 295 SE No.:_ 89-101 Source Documents PAP-0110, Rev. 3
' Description of Change This evaluation analysis changes made to PAP-0110, Shift Staffing and .
Overtime. Changes include personnel title changes'and increasing shift manning requirenients. i Summary I. No. This change is administrative in nature. It does not alter the
. plant design nor does it impact the accident analysis. Increasing the qualification requirement for on-shift personnel reduces the potential for personnel error. Therefore, it vill not increase the occurrence or the consequences of an accident previously evaluated in the USAR.
II. No. This change is administrative in nature and has no impact on accident analysis.- III. No. This change increases the qualification requirements for on-shift Auxiliary Operators and has no affect on Technical Specifications Table 6.2.2-1. I L
.F i
Attachment 3 L . PY-CEI/NRR-1141 L Page 182 of 285
-SE No.: 89-102 Source Document: DCP 87-0528, Rev. 2 Description of Change This design change replaced the Feedvater' Booster Pump Recirculation
- Valve IN27-F305 vith a CCI " Drag" valva. Additionally, a pair of flanges was removed and a 90' elbov was installed.
Summary I. No. The replacement of IN27-F305 with a CCI " DRAG" valve does'not alter-the function provided by the original valve. The " DRAG" valve meets the same codes as the original valve. Further, the new valve has bottom pressure characteristics which would reduce vibration and cavitation damage. The other piping modifications (use of an elbow and deletion of the flanges) are designed in accordance with ANSI B31.3 as was the original system design. Those modifications do-not alter the. function or operation of the Feedvater System Therefore, the probability or occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. Since all of the piping modifications made in this revision meet the same codes and standards used throughout the-Feedvater System and since these changes do not affect the function of the Feedvater-System, no new accidents or malfunctions are created. III. No. The Technical Specifications addresses the Feedvater Leakage Control System (3/4.6.l'.9) and containment isolation valve operability for valve, IN27-F599A/B (3/4.6.4). The piping modifications in this change are in the nonsafety-related portions of the Feedvater System and have no direct nor indirect affect on the safety-related-portions of the system nor affect the Feedvater Leakage Control-System or the N27 containment isolation valves.
p , > s - p Attcchm2nt 3 - PY-CEI/NRR-1141 L i Page 183 of 285 i V
- SE No.: 89-103 Source Document VLI-P50, Rev. 3-L Description of Change This change installs caps to prevent leakage from various vent and drain valves (not containment / boundary valves) of the containment Vessel Chilled Vater (P50) System. USAR Figure 9.4-22 is being revised to show this change.
Summary- - I. No. Installation of pipe caps increases the integrity.of P50. Pipe caps ' are not of significant mass, are not installed in areas of. safety-related equipment and cannot cause a failure of safety-related equipment. Therefore, this does not increase'the probability or consequences of an accident or malfunction of ; equipment. II. No. See to Item I above. III. No. This change is not considered in the bases of Technical Specifications, b , 4
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p h Att chm:nt 3 PY-CEI/NRR-ll41 L Page 184 of 285 L SE No.: 104
~ Source Document: DCP 89-0052, Rev. 0 Description of Change This design change makes structural modifications to the Transversing Incore Probe (TIP) drive platform to allow permanent removal of pool swell shield plates.
Summary I. No. The modifications in question reinforce the TIP drive support platform to allow the permanent removal of the pool svell shield , coverplates. .The TIP. drive units, which are nonsafety-related, must be protected from the pool swell event so as not to become a secondary missile.and impinge on safety-related equipment in the area. Per calculations (Reference File Code # 3:23.3) it has been demonstrated that the platform modifications, along with the two upper drive units being restrained, vill provide adequate means of protection and allow the platform to survive a pool swell event independently, without the vertical shield plates. Therefore, since the requirements set forth in USAR Sections 3.8.3, Appendix 3A and Appendix 3B are still met, the probability of the consequences from an accident previously evaluated are not increased by this change. II. No. 'The change in question results in an alternate method for protection of the TIP drive units from becoming secondary missiles in a pool ! swell event scenario. Due to the fact that the support platform and drive units remain protected from damage'due to pool swell loadings, the possibility for an accident or malfunction of a different type than those already evaluated is not created. III. No. .The change in question strengthens the support platform and restrains the two TIP drive units which sit on the platform to allow 1 all the drive units to survive a pool swell event. The actual operation of the TIP System is unaffected by these changes. Therefore, the safety margins on which the TIP System is based in the Technical Specifications remains unchanged. ; i t
Attachment 3 PY-CEI/NRR-1141 L Page 185 of 285 SE No.:- 89-105 Source Documents- DCP 88-0345, Rev. O Description of Change This design change adds an annunciator in the Control Room to alert the reactor operator when the reactor power goes below the Lov Power Setpoint (LPSP). Summary I. No. This design change adds nonsafety-related Control Room annunciation to altrt the operator that reactor power is at/below the Low Power Setpoiet. Per Technical Specification 3.1.4.2, the rod pattern controller must be verified operable following a power decrease below the LPSP.
-Addition of this. circuitry is not considered to impact the safety-function of the rod pattern controller as discussed in USAR Section 7.6.1.5 or Chapter 15. The design and function of the rod pattern controller is unaltered with respect to mitigating the
-consequences of the postulated control rod drop accident.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. See Item I above. III. No. Addition of the annunciator circuitry vill not change or affect the design / operability requirements for Technical Specification 3.3.6, Rod Block Instrumentation. As stated in Item I :5bove, this circuitry vill not affect the design / operability requirements for Technical' Specification 3.1.4.2, Rod Pattern Control system. Therefore, the margin of safety as defined in the bases for any Technical Specification is not reduced. e _ _ _ - - - _ _ _ . _ _ _ _ _ _ . _ _ -- - _ _ _ _ _ _ _ __-__________ ____________ _ _ ____ _
l Attachm nt 3 PY-CEI/NRR-1141 L Page 186 of 285 SE No.: 89-106-Source Document: DCP 88-0320, Rev. 0 Description of Change This change vill remove the Reactor Core Isolation Cooling (RCIC) and the Reactor Core Isolation Cooling / Residual Heat Removal (RHR) steam flow instrument-line. break trip unit. Trip signals from the Reactor-Isolation Cooling System isolation logic. However, these trip units vill provide an alarm function upon sensing an instrument line break. Summary I. No. The RCIC and RCIC/RHR instrument line break trip units monitor the RCIC and RCIC/RHR high flow transmitters for negative steam flov output, to determine if the transmitter impulse line has broken. The GE design specification data sheet states that thit monitoring function should result in an alarm output if a line break occurs. The elementary diagram and field viring show an RCIC isolation if an instrument line break occurs. The correct document is the GE design spec data sheets. .Since no credit for isolating RCIC on this instrument line break was taken in USAR Chapter 15,. deleting the isolation function from the instrument line break trip units vill not increase the probability of occurrence of an accident. II. No. The instrument line break instruments vill alarm if a break occurs.
.This alarm vill give the operators sufficient time to manually isolate RCIC if deemed necessary. Other leak detection instrumentation is used in the RCIC and RHR equipment areas to detect leakage, which will act as a backup to the flow monitors. If the flow monitor instrument lines break, and an alarm failure occurs, the steam output from the impulse line vill not significantly impact the surrounding environment. Therefore, no new accidents or malfunctions of equipment vill occur.
III. No. The instrument line break trip units are not included in Technical Specifications. Therefore, no margin of safety is reduced.
E t i Attochr:nt 3 1 PY-CEI/NRR-1141 L Page 187 of 285 SE No.: 89-107: Source Document DCP 87-0734A, Rev. 0-y Description of Change This design change modifies the Division 2 Diesel Generator (DG) pneumatic control / air supply systems for overspeed trip shutdown, emergency stop-shutdown, and normal engine shutdown on the engine. Summary I. No. These changes provide a rapid response overspeed trip shutdown system to protect the engine from possible damage in the event of
, governor failure at full fuel rack position. This design change '
retains the same basic on-engine functions and responses for the emergency stop shutdown and the normal engine shutdown portions of the system. The only significant difference resulting from these modifications-is the combining of the emergency stop signal with the overspeed trip signal. This signal is sensed by pressure switches in the P054D control panel as discussed and evaluated in the safety evaluation ~for DCP 87-0734B. The replacement components installed-as the result of this design change are of equivalent qualification and reliability as the original components. These design changes maintain the reliability of the overall l pneumatic shutdown system. They do not increase the probability of- ! L occurrence or the consequences of a malfunction of equipment because l the number of active inline, air supply, components is essentially I the same. II. No. All of the components installed as the result of this change meet all the requirements as demonstrated in Item I, above. The physical installation of the new components and their associated tubing i modifications are in accordance with the ' original installation requirements. This change creates no reductions of redundancy or. component performance levels when compared to the original design. [ Since no new component or design type is introduced by this change, l no new potential for equipment malfunctions is introduced by this i design change' document. ! III. No. The margin of safety as defined in the bases of Technical Specification Section 3/4.8 refers to the-reliability of the onsite i power supplies. As demonstrated in Items I and II above, reliability of the Division 2 Diesel Generator is not compromised by 1 this design change document. Ilence, the margins of safety described in the Technical Specifications bases is not affected. i
m Attachm:nt 3 PY-CEI/NRR-2141 L Page 188 of 202 SE No.: 89-108 Source Document: SOI N64/62, Rev. O Description of Change This safety' evaluation analyzes changing the Condenser Air Removal and the Offgas System operating instructions to allow the use of mechanical vacuum pumps during plant shutdown, and the use of offgas system purge (dilution) air to prevent the formation of hydrogen pockets during this L operation. The USAR currently states that the vacuum pumps are only used during startup to draw initial vacuum and that purge air is only used when preheating the recombiners and when air inleakage is less than 6 scfm. Summary i I. No. The mechanical vacuum pumps are currently evaluated for removing non-condensibles from the main and auxiliary condensers during plant startup. The formation of hydrogen from radiolytic decomposition of water in a reactor is proportional to reactor power, it is not or affected by the direction at which reactor power is changing but only by the reactor power at the time of formation. With reactor power decreasing during a shutdown, the amount of hydrogen produced per unit time is decreasing. Since at less than 5% reactor power, hydrogen formation is not sufficient to form an ignitable mixture of hydrogen in air in the mechanical vacuum compression stage during a startup, during shutdovn once reactor power is less than 5% hydrogen , formation is lov enough not to form an ignitable mixture in the ! vacuum pumps. The probability of occurrence or consequences of an ! accident is not increased by operating the mechanical vacuum pumps .; during plant startup and shutdown at less than 5% reactor power. ' Purge (dilution) air is currently evaluated for use when preheating . the recombiners and when air inleakage is less than 6 sefm. I i Dilution air is admitted to the Offgas Stream before the recombiners to ensure an adequate air supply to allow for recombination of the l radiolytic hydrogen. The admission of dilution air vill cause a lower probability of reaching a ignitable concentration of hydrogen l , ~in the Offgas System and. minimize the potential for forming i l ignitable hydrogen pockets. The probability of occurrence or a consequences of an accident is not increased by. admitting dilution ! air to prevent the formation of hydrogen pockets. As shown above, the probability of occurrence or consequences of an l accident is not increased. II. No. These' instruction changes do not alter the design function and operation of the N62 or N64 Systems. Therefore, the possibility for an accident or malfunction of a different type than any evaluated i previously in the USAR is not created. 2
Attachment 3 PY-CEI/NRR-1141 L Page 189 of-285 SE No.: '89-108 (Continued) Summary III. No. Operation of the mechanical vacuum pumps during plant shutdown or using dilution air to maintain the Offgas System free of hydrogen pockets'does not impact the bases for any Technical Specification. Therefore, the margin of safety as defined in the Technical Specification bases is not reduced. SE No.: 89-109 Source Documents- PNPP Emergency Plan, Rev. 9, TCN 2 Description of Change This evaluation analyzes various changes made to the PNPP Emergency Plan to ensure that the effectiveness of the plan has not been reduced, per 10 CFR 50.54(g), and to ensure that the plan continues to meet the standards of 10CFR50.47(b) and the requirements of Appendix E. Summary I. No. The Emergency Plan outlines the emergency preparedness actions and not the response of plant staff to an accident or equipment malfunction and, therefore, does not affect the probability of their occurrence. II. No. . The Emergency Plan does not direct the operation of plant systems or equipment and, therefore, does not create or increase the probability for an accident or malfunction. III. No. The Emergency Plan utilizes existing Technical Specifications and does not control'or effect their revision; therefore, the margin of safety as defined in Technical Specifications is not reduced. 1
. . . . . , . . . . . . . . . ~ . . . . . . ._... . . _ . . . . . . . _ . . . _ . _ _ _ _ _ _ . . . . - __ .
li E Attachttnt 3 PY-CEI/NRR-1141 L Page 190 of 285 SE No.: 89-110 Source-Document: DCP 88-0298, Rev. 0-Description of Change
.This design change removes the redundant 1E31-N351A-D Turbine Power i Complex Temperature Switches from MSIV isolation logic. It also moves the Turbine Power Complex temperature svitches alarm indicating function to their own alarm vindow.
Summary I. No. Temperature switches lE31-N351A-D, for the Turbine Power Complex vere added-by Gilbert Commonwealth to be redundant to the steam tunnel ambient temperature switches 1E31-N604A0A and delta temperature switches 1E31-N605A-F. The Turbine Power Complex switches operation, is not taken credit for in USAR Chapter 15, the GE Design Specification 22A3739AD or the plant Technical , Specifications. These temperature switches are located in a l nonsafety-related,'nonseismic category structure. Deleting the ! Turbine Power Complex Temperature Switches from the HSIV isolation I logic vill not increase the probability of the consequences of an l accident since the main steam line temperature switches N604/N605 ! (located in the same environmental zone) vill activate the isolation I logic if a steam leak occurs. The N604/N605-svitches are located in
'a seismic category I building and credit is taken for them operating, in the safety analysis, GE Design' Specification and Technical Specifications. j II. No. Since the Turbine Power Complex Temperature Switches are redundant ,
to the main steam line (MSL) temperature switches (N604/N605), their deletion vell not cause an accident. The MSL temperature switches will initiate the MSL isolation logic on a steam leak, making the isolation signal from the N351 temperature switches unnecessary. The N351 temperature. switches alarm function vill remain active to alarm on high temperatures in the turbine power complex. III. No. The Turbine Power Complex Temperature Switches are not included in the Technical Specifications.
rT - - Attcchnint 3 PY-CEI/NRR-ll41 L-Page-191 of 285
-SE No. / 89-111-Source Document: PAP-0202, Rev. 4, TCN4 Description of Change :
The revision to this procedure indicates the change of the offsite CEI telephone communication system from a_" Dimension" to a "Centrex" system.
- 7 Summary.
I. No. -The change of the Offsite Telephone System to the "Centrex" system T does not impact the ability to establish / maintain off-site . communications. The plant OPX network, which interfaces with the Centrex system is unchanged. Therefore, the probability of an accident or malfunction of equipment important to safety is not increased. II. No. The operation of.the OPX network within the plant remains the same. Therefore, the possibility of a new accident or malfunction is not created. III. No. The OPX network nor the offsite network have no interface with plant Technical _ Specifications. Therefore, the margin of safety regarding. Technical Specification bases is not reduced. SE No.: 89-112 Source Document: DCP 88-0314, Rev. O Description of Change This design change adds an alarm to the unit control console for the Reactor Recirculation (B33) System to alert the operator that the Automatic Flow Demand Limiter (AFDL) is in operation. Summary 1 i I. No. .The addition of an alarm to alert the operator when the AFDL is in operation does not alter the function or the operation of the B33 system. L II. No. New failure modes have not been created since the original design intent is still met. The AFDL alarm vill help the reactor operator better understand reactor recirculation operation. III. No. This item does-not affect the Technical Specifications, thus the margin of safety has not been reduced.
b:- Attcchzant'3
- f. PY-CEI/NRR-1141 L Page.192 of 285 SE No.: 89-113 Source Document: TAF-80559 Description of Change
, This evaluation analyzes the new P&ID 302-762 for the seal pressurization portion of the Penetration Pressurization and Personnel Air Lock Leakage control Systems (IP53).
6 . Summary c. I. No. The flex hoses, piping, tubing and supports are all ASME F Section III, Class 2 components,'where required, which is consistent vith present USAR commitments. The issue of this drawing has no impact on plant operation, analysis or design basis. Therefore, it. i does not increase the probability or consequences of an. accident or malfunction previously. evaluated or create the possibility of an accident or malfunction of a different type than previously evaluated. II. No. See item I above. III. No. The issue of this drawing does not affect the basis of any Technical Specification. i t i 1 l
t-
, Attachm:nt 3 H PY-CEI/NRR-1141 L Page 193 of 285 SE No.: 89-114 5 Source Document: DCP 89-0092, Rev. O i E
Description of Change This design change removes and plugs ASME Turbine Test lines (pressure tap and tracer injection) in the Extraction Steam (N36)-System since the ASME Turbine Test vill not be performed. Summary I. No. USAR Sections 10.1 and 10.2 do not reference the ASME test lines, which are planned to be removed and plugged. The ASME test was not performed and these lines are no longer needed. These test lines have no affect on safety-related equipment and'are not connected in any way to safety-related equipment. The ASME test line removal
~
vill not change the function or operation of the Extraction Steam System, the Condenser, and the Condenser Evacuation System addressed
'in USAR Sections 10.1, 10.4.1 and 10.4.2.- The deletion of these lines vill.not increase the potential for an accident of any type previously evaluated in the USAR.
.II. No. These lines have no interface with any safety-related system, and' were for testing purposes only. Therefore, this modification does not-create potential for a new accident.
III. No. The Technical Specification remains unchanged by this modification. This test is not identified in the Technical Specification, and the test lines do not affect the operational parameters of any equipment identified in the Technical Specification. _The margin.of safety, as . defined in the Technical Specification, is therefore not reduced. I i l' l
e Attechmsnt J PY-CEI/NRR-1141 L Page 194 of 285 SE No.: 89-115. Source Document: DCP 88-0339, Rev. O Description of' Change- I This design change replaces the' existing Offgas System (N64) loop seal - level switches with Drexelbrook point level control units. This vill e protect against loss of liquid and prevent gaseous escape to the drains. ! e Summary r
-I. No. This design replaces the instrumentation for monitoring vater level in.the N64 holdup line, prefilter and cooler condenser moisture
-separator loop-seals. The replacement instrumentation vill provide control to the automatic loop seal shutoff valve that prevents >
radioactive release from the system-in the event of a loss of loop seal level and also provides Control Room high and lov alarm " functions. The existing ITT Barton instrumentation vill remain installed but vill only be used for local indication which aids in
. loop seal filling process. This modification is consistent with the original design requirements.
The system analysis shows that the failure of the offgas loop seals vill not compromise any safety-related system or prevent safe shutdown. The new design with respect to connecting to the N64 System piping pressure boundary meets the requirements of the Standard Review Plan, Section 11.3, relative to design for hydrogen explosion limits. The Standard Review Plan places acceptance criteria on Regulatory Guide 1.143 (Design Guidance for Radioactive E Vaste Management Systems, Structures and Components Installed in . Light-Vater-Cooled Nuclear Power Plants) with respect to the use of ' threaded. pipe connections. Use of such connection in offgas piping is permissible for instrumentation. Such connections have been ' analyzed for hydrogen explosion and leakage, and will be verified by helium leak testing as specified in the Design Change work - instructions. Therefore, the probability of occurrence of an accident or malfunction of equipment.important to safety previously evaluated is not increased. II. No. The possibility for an accident or malfunction different than those previously evaluated has not been increased because the same design and installation parameters as the original design vere taken into account in the evaluation of this Design Change. III'. No. This Design Change vill not affect the margin of safety defined in the bases of .the Technical Specification Section 3/4 11.24 since the system function has not been altered. s
1 D' Attochment 3 PY-CEI/NRR-1141 L Page 195 of 285 SE No.: 89-116 Source Document: DCP 88-0234, Rev. 1 [ Description of Change The revision to this design changes valve 1P45-F713 from normally closed to locked closed and indicates additional testing requirements. r Summary. I. No. Valve IP45-F0713 vill be locked closed (LC) instead of normally closed (NC). The normally closed position was indicated in DCP 88-0234, Rev. O. The locked closed position vill further ensure that valve 1P45-F0713 is not inadvertently open, which could advctsely impact the ESV Loop C integtity. The EmergencyLService Water System (ESV) is classified as Safety Class 3 and Seismic Category I, with the primary safety function being to support the Emergency Core Cooling System. The 2-inch diameter fill line vill decrease the time required to fill Loop C of the ESV. This change does not alter the original function of ESV Loop C or the 3/4-inch keep fill line to Loop C as described in USAR Sections 9.2.1.1 and 9.2.1.2. The nonsafety portions of the 2-inch diameter quick fill line and 3/4-inch diamuter_ keep fill line vere leak tested per ANSI B31.1. The affected 3-inch diameter piping was tested in accordance with ASME Section XI, Article 1 VA-500. Therefore the alternate fill
- line addition vill not increase the probability of occurrence or the ,
consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR. i II. No. See Item I above. i III. No. The piping modifications covered in this DCP do not reduce the ! margin of safety as defined in the Technical Specification Sections: ? Tables T3.3.7.7-1 Remote Shutdown System Instrumentation of ESV ; Loops A & B, 3.7.1.1 ESV Limiting Conditions of Operation, and 3.8.1.1 ESV Electric Pover Systems. l v h v f f
4 Attechment 3 PY-CEI/NRR-1141 L Page 196 of 285 SE No.: 89-117
- , Source Document
- DCN 2000
- p. Description of Change This drawing change removes the Process Computer (C91) System points C11-NC068 and C11-NC069 from USAR Figures 4.6-5 and 7.7-1 since these 4
points are being removed from the computer data base. Summary I .. I. No. The Process Computer is a nonsafety-related system providing plant c information only, and has no control functions. All safety-related j;! signals fed to the Process Computer are optically isolated to assure j that operation of safety-related systems is not affected in any way p by the Process Computer System. Therefore, this change vill not h increase the probability of occurrence or the consequences of an ! accident. II. No. This change is a drawing change only. The function and operation of the Process Computer System has not changed. No new failure Modes have been created. III. No. The process computer is not addressed in the Technical Specifications, therefore vill not affect the bases of any technical specification. !~
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t.. Attochment 3 L PY-CEI/NRR-1141 L p Page 197 of 285 SE No.:' 89-118 p Source _ Document: DCN 2449 Description of Change
'This drawing change revises the Standby Diesel Generator Divisions 1, 2,
_ and 3 P&ID's to provide clarity for the interconnections between the-l various diesel and support systems and adds as-installed information that f was not previously shown on the drawings. Summary (f I. No. These drawing changes are editorial in nature and are being used to L clarify the drawings in question. The changes reflect the same h _ equipment configuration and safety classification as defined in USAR 4' Sections 9.5.4 through 9.5.8 and 9.5.9.3 for the Division 1, 2 and r 3 Standby Diesel Generators. Further, the function or operation of the Diesel Generators have not been altered. Therc!.,re, the i probability of occurrence or the consequence of an accident or malfunction of equipment important to safety previously evaluated is not increased, d g II. No. See Item I above. III. No. These changes are editorial only and are being uted to clarify the drawings. These changes do not alter the operation or reliability of:the Standby Diesel Generators. l: F
m i , Attechment 3 PY-CEI/NRR-1141 L Page 198 of 285 i SE No.: 89-119 Source Document: DCN 2673, Rev. O Description of Change This drawing change is editorial in nature. It corrects the labeling of
" locked closed" valves to " normal closed" valves on various fluid system drawings to eliminate inconsistencies in P&ID's, surveillance instructions, and valve lineup instructions. The P&ID's for the following systems are affected: Nuclear Boiler (B21), Reactor p Recirculation (B33), Leak Detection (E31), Reactor Core Isolation-Cooling (E51), Reactor Vater Cleanup (G33), Containment Vessel and Dryvell Purge (M14) Combustible Gas control (M51), Nuclear Closed i Cooling (P43), Instrument Air (P52), Penetration Pressurization (P53),
and Fire Protection (P54). Summary: I. No. Changes made to the figures within the USAR vill reflect the As-Built, As-Designed configuration and vill eliminate
, inconsistencies between the figures, the USAR text, the surveillance instruction utilized to administer operational state changes (evolutions). Elimination of these inconsistencies does not change the plant explicity or implicity. Elimination of inconsistencies in figurative depictions of the plant vill reduce the potential for error in interpreting As-Built, As-Designed configuration.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment 1. Sortant to safety previously evaluated in the USAR has not increated. II. No. -The possibility for an accident or malfunction of a different type than previously evaluated is not created by reducing the margin of g potential. interpretation error due to inconsistencies between L administrative procedures, and figurative depictions in the USAR. i III. No. This change-is. editorial in nature and does not change the plant as described in item I above. Since the operation of the systems in question is not altered the margin of safety as defined in the basis for Technical Specifications.is not reduced.- r I l.
s AttOchment 3 '; PY-CEI/NRR-1141 L Page 199 of 285 l i SE No.: 89-120- _ l Source Document: USAR CR 89-033 i CR 89-232 l l Description of Change ; This evaluation analyze the elimination of-the reference to "operctor j error" for a postulated Reactor Vater Cleanup System pipe break event l contained in USAR Section 6.2.1.1.4.2. . t Summary l
- 1. No. Evaluation of procedural initiation or erroneous operator initiation- d of containment spray for the postulated RVCU line break has no ;
effect on the accident evaluated.in USAR Section 6.2.1.1.4.2.- l Therefore, the probability of occurrence or the consequences'of i accident or malfunction of equipment important to safety is not ., reduced.- ' t II. No. See Item I above. j III. No. The change addresses a design evaluation related to containment I negative design pressure. Since the design evaluation has not . changed, the margin of safety as defined in the basis for Technical
- Specifications is not reduced. :
i t r i v p a i
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Attechment 3 .
- PY-CEI/NRR-ll41 L '
(f' Page 200 of 285 l
- SE No.: 89-121 l
Source Document: USAR CR 89-034 ,
.I Description of Change
'l This change is administrative and updates the USAR to reflect a i 1
reorganization of T21 at the corporate offices. i L t y Summary I. No. This change' alters-the' reporting relationship of CEI's home. office. However, the Vice President Nuclear Group,;onsite, stillLreports to [F the President-CEI at the home office. No physical change to the [ plant or to plant procedures are made, e j II. No. This is a change in the reporting relationship of CEI's home office. ic III. No. Technical Specification Section 6.2.1 remains satisfied by this organizational change. No other Technical. Specification or bases are affected. 1
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i r Attcchment 3 I L PY-CEI/NRR-ll41 L- , Page 201 of 285 SE No.: 89-122 Source Document: USAR CR 89-052 i Description of Change ' This change revises USAR Table 1.8-1 and Section 4.4.6.1.2 to state how the Loose Parts Monitoring System (LPMS) signal cables are routed from ' sensor to charge converter. Purther USAR Sections 4.4.6.1.2 and 4.4.6.1.4 vas revised to indicate that a manual calibration tool is used , i for calibration. i Summary. o
- 1. No. This change clarifies PNPP's installed field configuration for Loose Parts Monitoring System (LPMS) signal cabling. Additionally, PNPP ;
vill utilfre a manual calibration tool (pinger) procured from Combustion Engineering in lieu of the spring-loaded Starrett punch - for system calibration. This change doe not alter the function or , operation of the LPMS.
- As reflected in USAR 4.4.6.1.3, LPMS Safety Evaluation, the LPMS is i utilized for information purposes only by the Control Room operator. v The operator does not rely on such information for performance of ,
any safety-related action. Therefore, the probability of occurrence or the-consequences of any accident or malfunction of equipment ; important to. safety previously evaluated in the USAR has not been ! increased. II. No. LPMS operation is not altered. No other systems or components are i affected by the operation of LPMS. Therefore,.the possibility of an accident previously evaluated in the USAR has not been created. III. No. This change.does not impact the LPMS operability and surveillance requirements stated in Technical Specification 3/4.3.7.8 nor the . description in the associated Technical Specification base. ! Therefore, the margin of safety defined in Technical Specifications
-is not reduced.
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- Attcchment 3 PY-CEI/NRR-1141 L Page 202 of 285 SE No.: 89-123 Source Document: USAR CR 89-057 Description of Change The change returns the Post-LOCA minimum suppression pool vent submergence to its original design value of 2'0". '
Summary I. No. This change returns suppression pool vent submergence to its original design basis value. Actual suppression pool vent coverage does not affect the probability of occurrence of an accident. Returning the vent submergence to its original design minimum value vill ensure that a water seal vill be maintained for the RHR A test return line for a 30 day period following postulated accident conditions. The assurance of a water seal vill eliminate a potential secondary bypass leakage path. All other lines which take similar credit for a water seal are not adversely impacted by this , change. Return of the Post-LOCA minimum suppression pool vent submergence to its original design value, provides an increase in the plant safety margin. Previous analyses verified that-less vent coverage provided sufficient coverage to ensure complete steam condensation under , design basis conditions. The increased level vill provide additional design margin. Therefore, the probability of occurrence or conseq>ance of an accident or malfunction of equipment important ' to safety is no increased. II. No. Notation of actual calculated minimum vent coverage does not deal with any new accident or malfunction. III. No. Testing has shoved that complete steam condensation vill occur with vent coverage of l'8-15/16". Calculated minimum long term post-accident suppression pool water level exceeds l'8-15/16" by 3'. Therefore, the margin of safety remains unchanged.
i AttOchment 3 . PY-CEI/NRR-1141 L 4 Page 203 of 285 ; SE No.: 89-124
- Source Document: MFI 1-89-124 Description of Change This safety evaluation analyzes the installation of Emergency Closed ;
Cooling System Pump Notor 2P42-C001B, in the Control Room Emergency Recirculation System as a temporary replacement for motor OM26-C001A. .
- Summary !
n o 1. No. The temporary motor is a calculated 2 KV larger load than the original motor evaluated in Table 8.3-1 for Unit 1. Division 1 Diesel. The calculation as documented in Calculation WHISC-0004, identifies.the temporary motor as having a 1 ampere lover full load : current, but with a higher-power factor. The end result would be an ! 82.0 KV load, instead of the original 80.0 KV load. This additional 2'KV load is an insignificant addition to the maximum allowable
,7,000 KV-Unit 1, Division 1 load. Therefore, the probability of ,
occurrence or consequence or an accident of malfunction of equipment ! is not increased. II. No. The. modified motor has a larger cervice factor, thus has a greater motor operating safety range than the original motor. Thus a new - malfunction or accident is not created. I III. No. The 1 ampere reduction in the full load ampere is within the ! postulated diesel load " band" described within Technical
- Specification 3/4.8.1.1.2 for Unit 1, Division 1 loads. Therefore, e a margin of safety is not reduced. '
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1 Atttchmen1 3 1 [ PY-CEI/NRR-ll41 L j Page 204 of 285 SE No.: 89-125 ; Source Document: DCP 89-0088, Rev. O Description of Change This design change installs security bars on the four Reactor Building
- Containment Vacuum Relief (M17) System breaker lines. This vin improve
. plant security by ensuring there vould be no access via the M17 pipe from , a protected area into a vital area. Summary , I. No. Containment vacuum breakers are required to prevent the containment atmosphere reaching 0,8 psi negative design pressure. There are four vacuum breaker lines. Two lines are more than adequate to break vacuum and maintain containment integrity. Based upon the revision of calculations M17-10 " sizing of containment vacuum relief valves" and calculation 2.5.3 " containment vacuum relief analyses" changes in the flow resistance coefficient due to the addition of the security bars did not change the results - of the calculations. Therefore, the two out of four containment vacuum breaker penetrations can still pass the required flow to be ; within the previously calculated maximum containment negative pressure of 0.72 psi. Thus, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. See item I above. III. No. Since the calculated maximum negative pressure for the containment did not change, the margin of safety between original calculation and the negative pressure design basis of the containment remains the same. Thus, the margin of safety is not affected.
e Attachment 3 PY-CEI/NRR-ll41 L ' Page 205 of 285
. SE No.: 89-126 Source Document: USAR CR 89-064 CR 89-216 i
Description of Change This change modifies USAR Table 6.2-40, Notes 3 and 5 to permit reverse , testing of valve IN27-F751. Summary I. No. A test pressure of 15 psi or less does not provide a sufficient
- force to alter the seating characteristics'of the valve. Thus, the test pressure applied to either the inlet or the outlet of the valve vill provide equivalent results. Section XI of ASME Boiler and Pressure Vessel Code " Rules for Inservice Inspection of Nuclear
- Power Plant Components" paragraph IVV-3423(d), specifically permits
. differential pressure testing in this menner, for valves (except check valves) if the function differential pressure is 15 psi-or ,
less. This change is consistent with the requirements of 10 CFR 50 t Appendix J for Type C test since, as stated above, equivalent results vill be obtained. Therefore, the probability of occurrence ' or the consequences of an accident or malfunction of equipment is ; not increased, i II. No. Based upon Item I above, it can be concluded that an acceptable i Local Leak Rate Test (LLRT) in the direction opposite normal flow, where the functional differential pressure is 15 psi or less will - repeat itself when the flow is in'the normal direction. Therefore, no new malfunction is created, t III. No. Based upon Item I above, operability of the tested valve is not affected. Leakage criteria as stated in the bases of the Technical ! Specifications are not changed. i i b P i E e
Attcchment 3. PY-CEI/NRR-1141 L
-O Page 206 of 285 SE No.s. 127 Source Document: SCR 1-88-1482 through 1-88-1485 Description of Change This setpoint change lovers the Turbine Power Complex Temperature Switches 1E31NO351A-D to reflect alarm settings rather that isolation settings.
Summary .
.I. No. The Turbine Power Complex Temperature Switches vill no longer initiate a main steam.line iso.lation. The switches only provide a high temperature alarm function (Reference DCP.88-0298). The new alarm setpoint .is set'below the old isolation setpoint to provide earlier warning of a steam leak in the steam-tunnel. Having only an alarm setpoint vill not increase the probability of the consequences of an accident since the main steam line temperature switches N604/N605 vill activate the isolation logic if a steam leak did occur.
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- 11. No. Turbine Power Complex Temperature Switches will not initiate the main steam line' isolation logic on a steam leak, making their isolation setpoint unnecessary. The revised setpoint vill not_cause a new accident, or malfunction of equipment since itr only function is to alarm.
III. No. The Turbine Power Complex Temperature Switch setpoints are not part
- of the. Technical Specification requirements. and vill not decrease - ,
the margin of safety as described in the bases of the Technical -
-Specifications. i t
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Attcchment 3 l PY-CEI/NRR-ll41 L f Page 207 of 285 SE No.: 89-128
. Source Document: DCP 87-0300, Rev. O !
Description of Change. ! This design change installs duct vork to supply air from the Turbine ! Building ventilation (M35) System to the Emergency Response Information
- System (ERIS-C95) Data Acquisition System (DAS) cabinets IH22-P110C-1 and
- 2. located on the turbine deck for cooling.
I Summary I. No. This design change revises the supply air distribution system within-the Turbine Building. A new branch is being added to provide cooling.to ERIS panels to permit acceptable ERIS performance during all operating conditions. M35 System operation has not been changed 3 nor has the total system flow rate be altered. Based upon the fact that supply air to the ERIS panels vill provide more reliable ERIS ; service and overall H35 System function has not changed, the parameters upon which the accident analysis in the USAR is based, have not been affected. . II. No. This change vill provide better supply air distribution without the overall system. design / function being affected. Therefore, l malfunctions of a different type vill not be created. III. No. The Technical Specifications do not address the Turbine Building Ventilation System or ERIS. The change to the supply air distribution system does not affect M35 System operation. This does . increase reliability of the ERIS panels. Therefore, no margin of safety is affected. C b 4 1 I F I I'
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?re - Attachment 3 PY-CEI/NRR-1141 L Page 208 of 285 SE No.: 89-129 Source Document -MFI 1-89-129 Description of Change f This change relocates the sample cell flow orifices in the ! 1H51-P022A/B H, Analyzers to a point downstream of the sample flow meters ! and installs a8justable metering valves in the orifices original location. ! Summary i
- 1. No. This change does not change the function or operation of the H Analyzers. The modification improves the ability to maintain '
H Analyzer calibration and operability. Therefore, the capability ; t monitor H 2. concentration levels after an accident is improved. Incorporating this change vill not effect the operation of the flov _ switch which monitors for lov flov through the analyzer cell. The ! switch vill still function to provide a " low flow" alarm as- i described in USAR Section 7.3.1.1.11. Therefore, the probability of , occurrence or consequences of an accident or malfunction of . equipment important to safety is not reduced. II. No.--See Item I above. - III. No. This function and operation of the H analyzer 2 has not changed per l , Technical Specification 3.3.7.5. Therefore, the margin of safety ! has been improved.= ! v s t i e b
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Attcchment 3 l PY-CEI/NRR-1141 L ' Page 209 of 285 ; SE No.: 89-130 89-188 Source Document: NR PPDN 1336, Rev. O NR PPDN 1336, Rev. 1 ' Description of Change Hydrolasing of the High Pressure (HP) Condenser has caused the condenser , tubes to deform vithin the tube sheet. Approximately 120 tubes are affected and must be rerolled or plugged. - Safety Evaluation 89-188 corrects the heat transfer rate decrease percentage specified in Safety Evaluation 89-130 and notes that rerolling satisfactorily solved this problem. Summary I. No. The repairs vill either return the system to its original operating condition by rerolling the tubes or to operation with a slightly lowered heat transfer rate (0.3%) by plugging the tubes. This repair vill not alter the design of the Circulating Vater System with respect to the prevention of injection of radioactive material into the circulating water with its subsequent release to the atmosphere through evaporation in the cooling tover (reference USAR-Section 10.5.3). Further, llSAR Section 10.4.1 states that the Circulating Vater System providos no safety-related function. Therefore, this repair vill not increase the probability of occurrence or the consequences of the accident or malfunction of equipment previously evaluated in the USAR. II. No. The plugging of the tubes does not create the potential for a new accident as.the only aspect of condenser operation affected is the overall heat transfer rate in the H.P. condenser. This heat transfer rate decrease is at most 0.3%, assuming the condenser tubes vould be plugged from this repair. Potential leakage through the condenser has been previously evaluated in the USAR and is not i affected by this repair. System analysis has shown that failure of , the Circulating Vater System vill not compromise any safety-related system or prevent safe shutdovn. Therefore, no potential is created t for an accident not previously evaluated in the USAR from this repair. III. No. The number of tubes in the condenser or the condenser heat transfer rate is not addressed in the Technical Specifications. The repair of the tubes vould not create the possibility of condensate leaking into the circulating water. Ensuring there is no uncontrolled release path to the environment.
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AttOchment 3 ' PY-CEI/NRR-1141 L : Page 210 of 285 ,
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- SE No.: 89-130 (Continued) l r .! , 89-188 i Summary
[ i i Any. leakage of circulating vater into condensate vould be identified , by conductivity elements in the condenser-(USAR Section 10.4.1.4) r aiding this operator in maintaining vater chemistry limits. " Therefore,.the margin of safety in the Technical Specifications is not affected or reduced. , f r
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Attachment 3 PY-CEI/NRR-1141 L Page 211 of 285 [ SE No.: 89-131 Source Document: LL & JED's 1-89-130, 1-89-131 l Description of Change: { These temporary modifications disable the Division 1 and Division 2 ; Diesel Generator (DG) neutral overvoltage relays IR2200607 and 1R2200703. : Summary: ! I. No. The neutral overvoltage trips are disabled in the event that the DG- , : has started due to a loss of coolant accident:(USAR Section 8.0). This temporary modification disables the neutral overvoltage tript ; during all modes of DG operation. Further, the alarm functions ' associated with these relays vill be disabled. Adequate compensatory administrative actions shall be taken to monitor . generator neutral voltage during DG operation. The probability of a ground on safety-related equipment occurring is not increased by , removing the generator neutral trip / alarm function. ' II. No. Disabling the subject relays does not create a possibility for a malfunction or accident which is outside of the USAR evaluation. i The Diesel Generator operates in response to a LOCA or a Loss of Off-site Pover. The malfunction of concern, a ground fault, cannot be caused by removing the trip function. The consequ<nces of the
- ground fault are enveloped by the USAR LOCA Analysis. i III. No. The subject relays are not specifically discussed in the Technical !
Specifications. The Technical Specification bases for electrical power sources are to ensure adequate power availability to supply , safety-related equipment for safe shutdown and accident mitigation. '- This temporary modification vill not affect these bases. Therefore, safety is not reduced. ;
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p ~ w Attechznt 3 PY-CEI/NRR-1141 L ! Page 212 of 285 e l SE No.: 89-132 i Source Document DCP 87-0800, Rev. 2 Description of Change 4 This design change replaces the existing 1/2-inch turbine meter (ITT ; Barton) vith a 1-inch magnetic flovmeter (Roccaount) due to a rust ' and fiber coating being found on the internals of the original meter > (IE31-H021).
- i Summary j c I. No. Flov element (1E31-N021) and associated transmitter (1E31-N0695) are nonsafety-related, Seismic Category I. The system for monitoring Reactor Coolant Pressure Boundary leakage in the dryvell is not part of the plant protection or safety system. USAR Sections 5.2.5.2.11 Detection of Leakage within Dryvell, 5.2.5.2.1.c, cooler condensate .!
Drains 5.2.5.10, Regulatory Guide 1.45 Compliance; and 9.4.6.2.1,
.Dryvell Cooling Systems are not affected by this change since only the method of detection has changed (turbine meter versus magnetic '
flow meter). Therefore, the probability of occurrence or the , consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. l
- 11. No. Testing of the condensate from the Dryvell coolers indicates that the conductivity of this vater is of a level which would ensure ;
satisfactory flovmeter operation. . Functional Test, TAF80297, confirms that any potential biological - growth inside the flovmeter vill not affect the meter's output ; signal. ' A " Qualified Life Test," Report A))-03479, was conducted based upon the monitoring of construction. A qualified life of ten years was determined for the flovmeter when subjected to a' harsh environment - (DV-1). A functional test, Test Procedure / Report DCP87-00800, vas performed following a seismic test found that the flow components are qualified for the intended function of monitoring condensate flov. ' Therefore, the possibility of an accident or malfunction of'a different type than evaluated is not created. III. No. Technical Specification 3.4.3.2 identifies only total allovable leakage. Condensate flow rates from the Dryvell Coolers has not changed. The Leak Detection Systems specified in Technical ! Specifications 3/4.4-9 and 3.4.3.1 have not been altered. Therefore, the margin of safety is defined in the bases for these - Technical Specifications have not been reduced. l l
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, .Att0chment 3 ..
j
.PY-CEI/NRR-1141'L- !
Page 213 of 285 l
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~ SE No.: 89-133: .!
Source Document: DCN 2732 '! oeserietion or Channe- lt This drawing change corrects the design and operating data on USAR Figure ll
+. 7.3-10 (P&ID D-912-606). :
Summary i I. No. . Revision to'the P&ID corrects the temperature and pressure data for- :
. the flov paths shown on the drawing. The design' data which the USAR. !
analyses were based remains the same (Reference Circulation M16-3). ;
i Thus, there is no affect on the equipment or to the analysis'in the USAR. "
*t II. No. The design basis data remains the same as originally analyzed in the a '
USAR.
. Thus, no new accident / malfunction is created.
III. No.- The design basis does not change. Thus, the margin of safety-is not '5
-affected. i
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t Att:ch cnt 3 PY-CEI/NRR-1141 L Page 214 of 295 SE No.: 89-135 Source Document: DCP 88-072B, Rev. O Description of Change This jesign change modifies the Division 3 Diesel Generator (DG) neutral ground overvoltage relay 1R22-00803 (Device 59NG) from a trip / alarm device to an alarm only device. Summary I. No. The DG neutral ground overvoltage relay trip function is bypassed during a LOCA vith the alarm function being maintained (USAR Section 8.0). This Design Change vill modify the 59NG device to be alarm only for all modes of DG operation. Administrative actions similar to those presently taken during a LOCA should a ground fault occur shall now be taken in all modes of DG operation. The probability of a ground fault on safety-related equipment occurring is not increased by removing 59NG's trip function. II. No. Hodifying the Device 59NG vill not create the possibility for an accident or malfunction which is outside of the USAR evaluation. The DG operates in response to a LOCA or to a Loss of Offsite Power. The malfunction of concern, a ground fault, vill not be caused by modifying the relay to be alarm only device. The consequences of the occurrence of a ground fault are enveloped by the USAR LOCA analysis. 4 III. No. The device 59NG relay are not specifically discussed in the Technical Specifications. The Tcchnical Specification bases for electrical power sources are to ensure adequate power availability to supply safety-related equipment for safe shutdown and accident mitigation. This modification vill not affect these bases. Therefore, the margin of safety is not reduced. l l
iM Attechment 3 i PY-CEI/NRR-ll41 L i- Page 215 of 285
- SE No.: 89-136 Source Document: DCP 87-0276E, Rev. O Description of Change ,
This design change modifies the Division 1 and Division 2 Diesel Generator (DG) neutral ground overvoltage relays 1R22-00607 and IR22-00703 (Device 59NG) from a trip / alarm device to an alarm only device.
- Summary I. No. The DG neutral ground overvoltage relay trip function for each DG is i bypassed during a LOCA vith the alarm function being maintained (USAR Section 8.0). This Design Change vill modify the 59NG device for each DG to be alarm only for all modes of DG operation.
Administrative actions similar to those presently taken during a LOCA should a ground fault occut shall nov be taken in all modes of - DG operation. The probability of a ground fault on safety-related equipment occurring is not increased by removing 59NG's trip function. II. No. Modifying the 59NG devices vill not create the possibility for an accident or malfunction which is outside of the USAR evaluation. . The Division 1 and 2 Diesel Generators operate in response to a LOCA or to a Loss of Offsite Pover. The malfunction of concern, a ground fault, vill not be the result of modifying the relays to be , alarm only devices. The consequences of the occurrence of a ground
- fault are enveloped by the USAR LOCA analysis.
III. No. The device 59NG relays are not specifically discussed in the Technical Specifications. The Technical Specification bases for electrical power sources are to ensure adequate power availability to supply safety-related equipment for safe shutdown and accident mitigation. This modification vill not affect those bases. Therefore, the margin of safety is not reduced. 1 1.
Attcchment 3 PY-CEI/NRR-1141 L Page 216 of 285 l SE No.: 89-137 Source Doccment: DCP 88-0220, Rev. 0 l l Description of Change This design ~ change upgrades the Loose Parts Monitoring (LPMS-R63) System by changing the frequency range monitored by the alate bistable to a range that is more conducive for detecting loose parts. This change also , installs an improved monitoring / digital recording system for ease of analysis and alarm data discrimination. i Summary I. No. This change does not alter the function or operation of the Loose l Parts Monitoring System. This system is used for information ' purposes only by the plant operator. The operator does not rely on , the information provided by the loose parts monitor for the i performance of any safety-related action. Therefore, the : probability of occurrence or possibility of malfunction of equipment important to safety previously evaluated in the USAR is not i increased. : II. No. This change does not alter the function or operation of the Loose ' Parts Monitoring System.- This system is used for information I purposes only by the plant operator. _Therefore,-the possibility for an accident or malfunction of a different type than any previously ! evaluated in the USAR is not created. III. No. Since this change does not alter the function or operation of the LPMS the margin of safety as defined in the bases for any Technical Specification is not reduced. b 6 I - r
Attcchment 3 PY-CEI/NRR-1141 L l Page 217 of 285 i SE No. 89-138 Source Document: MFI 1-89-138 ' Description of Change !
. Temporary air connections are requited to maintain instrument air -
services in containment while performing Iristrument Air (PS2) System modifications. 3 l Summary I. No. This modification simply reroutes air from connections located , outside of the containment.through temporary hoses to connections ; inside of the containment to maintain service to containment end-users. The PS2 System has no safety-related functions.except for maintaining containment integrity (USAR section 3.2). the analysis for a loss of instrument air contained in USAR
-Section 15.2.10 is applicable during " Normal" Plant Operation. ,
Currently the Plant is shutdown, in mode 5, with containment a integrity not required. Therefore, the probability of occurrence of ' the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not ., increased. II. No. See Item I above. III. No. Temporary air jumpers do not affect any Technical Specifications since installation and use of jumpers vill only occur when containment integrity in not needed. The requirements of the action statement for Technical Specification 3.6.1.1.2, Primary Containment-Integrity, vill be met at the time of this MFI installation /use. Therefore, no margin of safety as defined in the bases for any : Technical Specification is affected. s
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Attcchr:nts3 ' PY-CEI/NRR-1141 L Page 218 of 285 SE No.: 89-139 Source Document: USAR CR 89-068 , Description of Changes i >+ This evaluation analyzes changes to USAR Section 13.1. The Title " Senior- ! Operations Coordinator" vas changed to " Superintendent,. Plant operations." Table 13.1-1 was changed to indicate that the Manager, j ' Operations Section is the ANSI 18.1-1971 cequired Operations Manager. ; Further the Superintendent,' Plant Operations is qualified to the i requirements of.the ANSI 18.1-1971 Operations-Manager position. { Summary: i I . -; No . These changes are simply title _ changes. The people filling these . positions are'the same individuals who~ vere filling the positions ! prior to the title changes. ! II. No.' These changes do not affect any plant system, structure, or j component. ~
.i III. No. 'The Technical Specification bases are not affected by these title i , changes. !
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i L AttOchment 3 L PY-CEI/NRR-1141 L
! Page 219 of 285 f
E SE No.: 89-140 Source Document: TXI-084 Description of Change This temporary instruction allows the withdrawing of a control rod by .; using the electrical override method. Summary l-
! I. No. The Control Rod Drive Hydraulic (CRDH) System is important to safety. TXI-084 prevents a malfunction of this equipment by placing the plant equipment in a safe condition and using a temporary solenoid coil to perform the actuation of a directional control valve which cannot be actuated in the manner required using the Rod i Control and Information System.
i The actual method that TXI-084 is based upon is described in the Operation and Maintenance Instructions for the control Rod Drive System, GEK-75598B, Section 3.35. TXI-084 also allovs an increase of drive water pressure above that allowed by S01-C11 (CRDH). Drive water pressure is still limited below $50 psi by the in-line relief valve IC11-F040. The in-line relief valve is provided to avoid exposing a drive to a pressure that vill cause the drive to move in excess of the allowable drive speed. There are several Rod Vithdrawal Error accidents described in USAR Section 15.4. The only one applicable to the plant conditions during which TXI-084 is performed is Section 15.4.1.1, Control Rod Removal Error During Refueling. TXI-084 does not disable or bypass any of the interlocks which provide assurance that inadvertent criticality does not occur because a control rod has been removed or is withdrawn in coincidence with another control rod. To minimize the possibility of loading fuel into a cell containing no control rod, it is required that all control rods are fully inserted when fuel is being loaded into the core. This requirement is. backed up by refueling interlocks on the refueling platform. When the mode switch is in the " refuel" position, the interlocks prevent the platform from being moved over the core if a control rod l is withdrawn and fuel is on the hoist. Likevise, if the refueling platform is over the core and fuel is on the hoist, control rod motion is blocked by the interlocks. TXI-084 uses a temporary solenoid coil only on the insert directional control valve EP-123. The withdrav directional control valve EP-122 vill continue to be actuated by the Rod Control And Information System. In this configuration control rod withdrawal is still subject to all the interlocks of RCIS. l i
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Attcchment 3 i
-PY-CEI/NRR-1141 L Page 220 of 285 l SE No.: 89-140 (Continued)
Summary : L Vhen the platform is not over the core (or fuel is not on the hoist) i L h and the mode switch is in the " refuel" position, only one control rod can be withdrawn. An attempt to withdraw a second rod results in a rod block by the refueling interlocks. Since the core is designed to meet shutdown requirements with the highest worth rod , withdrawn, the core remains suberitical even with one rod withdrawn.
.TXI-084 does not disable or bypass any of the interlocks designed to I prevent withdrawal of a second rod.
L II. No. Control rod withdrawal has been previously evaluated in the USAR. None of the changes made by TXI-084 creates an accident or j malfunction of a different type than has been previously evaluated ; . in the USAR.- 4' E l.. III. No. Technical Specifications applicable to refuel mode maintenance or - repair of control rod drives ensure that such maintenance vill be i performed under conditions that limit the probability of inadvertent; criticality.'-As discussed in Item I, TXI-084 does not disable or ; bypass any interlock designed to prevent a second control rod from ' being withdrawn from the core.. Therefore the margin of safety has ! not been reduced. I
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s L Attachment 3 l PY-CEI/NRR-1141 L page 221 of_285 ; SE No.: 89-141 Source Document: ONI-ZZZ-2, Rev. I j f USAR 89-155
- , Description of Change I
- This safety evaluation concerns the se-analysis of an accidental release of natural gas-from a break in the closest 4-inch natural gas pipeline.
This re-analysis involves the identification of any differences resulting ; from an increase in line pressure in the 4-inch pipeline from the ' original 35 psi to a pressure of 60 psi and the change in pipe depth from ; the original 32-inch depth to a 30-inch depth. . ' Summary i k I' No. The consequences of a natural gas pipeline rupture vere previously evaluated in USAR section 2.2.3.1.1.2. The new analysis shows that [' the increase in pipeline pressure from 35 to 60 psi (for the most limiting pipe - 4" line closest to the plant) vill not effect the , I original conclusions in USAR section 2.2.3.1.1.2 (Reference i Analysis 058U-R&RA-A1). Therefore,-the probability of the '
- . occurrence or consequences of an accident is not increased.
i II. No. See item I above. III. No. This change does not affect Technical Specifications. b b f k i l ; t
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Attcchm:nt 3 PY-CEI/NRR-1141 L Page 222 of 285 SE No.: 89-142 Source Document: DCP 89-0117, Rev. 0
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i Description of Change This design change adds a 30 second time delay to the Control Complex Chillers (P47) LOOP /LOCA initiation circuitry to allow the Nuclear Closed i Cooling (NCC) valves to close, Emergency Closed Cooling (ECC) valves to open and flow switches OP42-N306A/B to reset for an auto-start of the chiller. Summary I. No. This design change installs a 30 second time delay on the Control Complex Chiller circuitry during a LOOP /LOCA initiation. This will provide sufficient time for the NCC to ECC chiller cooling water change over which allows reset of flow switches OP42-N306A/B. A 30 second delay following a LOOP /LOCA initiation does not affect control room temperature and vill assure an automatic start of the chillers. The control logic change simplifies the circuitry by deleting I control relay from each division. The Control Complex Chillers are the last major load to be sequenced onto the diesels so the diesel KV loading calculation (26133-86-2) is not adversely affected nor is the voltage profile. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety r eviously evaluated in the USAR is not increased. II. No. See Item I, above.
.III. No. The auto initiate timing sequence of the Control Complex Chillers is not described in the Technical Specifications. Therefore, the margin of safety as defined in the basis for Technical Specifications is not reduced.
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l w- Attcchm:nt 3 PY-CEI/NRR-1141 L i: Page 223 of 285 it SE No.: 89-143 J Source Document: DCP 89-0098 I Description of Change This design change adds vent lines to the scram discharge volume (SDV) header level transmittete. This vill allow the impulse lines to properly drain after the SDV drains. Summary I. No. The subject change adds vents to the current SDV Level instrument lines. Those vents do not alter the function or operation of the SDV Level instrument or of the SDV. Due to the redundancy of the SDV Level instrumentation, this modification does not change the 1 probability of common mode failure nor does it alter the single failure criteria. Further, this. change does not impact the Standby Liquid Control (C41) System. The C41 system provides an alternate shutdown capability to the control rods. Therefore, the probability < of occurrence or the consequences of an accident or malfunction of equipment importance to safety previously evaluated is not increased. II. No. Failure of.one or more of the newly added vent lines is bounded by an instrument pipe Ifne break which less conservatively considers breakage with a direct leakage path for the reactor coolant inventory. Failure'of the SVD Level instruments results in conservatively shutting down the plant (automatically). Hence, the possibility of a new accident or malfunction not previously evaluated is not created. III. No. Technical Specifications Sections 3/4.1.3, 3/4.3.1 and 3/4.3.6, including their bases, are unaffected by this change. Therefore, the margin of safety contained in these Technical Specifications is not reduced. t t l
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.c SE No.: 89-0144 Source Document: TDI DR/OR Report, Revision 1, Dated February, 1985
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Description of Changes *
'This evaluation provides justification for not implementing certain TDI DR/0R modifications committed to in the TDI DR/0R Report, Revision 1, dated February 1985.
Summary: v
.I. No. These deviations were all evaluated against original TDI Owners Group criteria and determined to be acceptable in their present configurations. There is no increase in the probability of occurrence or the consequences of an accident or malfunction of w equipment previously evaluated in the USAR.
II. No. No additional changes were made to the diesel generators; therefore, the possibility for an accident or malfunction of a different type than any previously evaluated is not increased. These deviation do not change the reliability of the diesel'
'III. No.
generators; therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced. E rx
pJ p-Att:chmsnt 3 PY-CEI/NRR-1141 L b Page 225 of 285 L "SE N5.: 89-145 t Source Document: PAP-1920, Rev. 3 i Description of Change This evaluation examines changes'to PAP-1920, Periodic Fire Inspection. Changes included trending satisfactory conditions to develop historical- - data and improved methodology for conducting. periodic inspections. Summary
)
See Safety Evaluation 88-470. SE No. ' 89-146 Source Document: SXI-040 i
. Description of Change This special test instruction redistributes the Turbine Power Complex Ventilation (H42) System airflows at Elevation 647-O' of the Turbine Power Complex to provide additional cooling for the motor control centers.
Summary I. No. USAR Section 9.4.8.3, states that the operation of the Turbine Power Complex Ventilation System is not required for the safe shutdown of the plant. Further, there are no divisional power supplies located on the'647'-0" Elevation. Therefore, the probability of' occurrence or consequence of an accident or malfunction of equipment is not increased. II. No. 'See Item I above. ' l-L III. No. 'The Turbine Power Complex' Ventilation System is not required for safe shutdown of the plant.
l u Att:chr;nt 3 PY-CEI/NRR-ll41 L Page 226 of 285 SE No.:' 89-147 ' Source Document: DCP 87-456Y, Rev.
Description:
This Design change evaluates the architectural and structural changes made to the Service Building Machine Shop to convert it into a radiological acceptance hot machine shop.
-Summary:
- 1. No. The Service Building is not a safety-related structure, and as such any architectural changes made to thf. facility do not-involve a change to safety-related structures or equipment. The addition to '
the existing facility is nonsafety-related. The addition is located between the Service and the Intermediate Buildings. The addition is a_ free standing structure, designed to meet the specifications defined for the existing Service Building. Therefore, the probability of an accident / malfunction of equipment important to safety is not increased. II. No. The addition was designed for the applicable design load conditions (including tornado wind) and combinat. ions, therefore the possibility of an accident / malfunction of a different type is not created. III. No. The changes presented here do not involve changes to any bases for Technical Specifications. Therefore, no margins of safety have been reduced. i 1 7
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Page 227 of 285 SE NO. 89-148
, SOURCE DOCUMENT: PAP-1917, REV. 2, TCN 2 DESCRIPTION OF CHANGE This evaluation analyzes a change made to PAP-1917, Fire Protection Training Programs. The change is concerned with the requalification ,
training of Fire Brigade Personnel. Summary I. No. All changes made to this procedure vere evaluated in respect to the USAR. All changes made vere found to be consistent with the fire protection requirements of the USAR. Further,-the USAR and 10 CFR.50 Appendix R do not specify the content of Fire Brigade Requalification Training Programs. Therefore, the probability of an E3 occurrence or consequence of an accident previously evaluated in the
- USAR is not increased.
II. No. All changes made to this procedure vere evaluated and found to be consistent with the fire protection requirements of the USAR. Therefore, the possibility of an accident different from any previously evaluates in the USAR has not been created.
- III. No.- The Fire Protection Program is referenced in sections 6.5.1.6, 6.5.2.8, and 6.8.1 of the Technical Specifications. This evaluation 1 has determined that all changes made to this procedure are consistent with these sections of the Technical specifications.
Therefore, no margin of safety has been reduced. , L-
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'SE No.: 89-150 l Source Document: DCN 2773 Description of Change
,v This drawing change involves updating the Turbine Building Chilled Vater (P46) System,.the Control Complex Chilled Vater (P47) System and the Containment Vessel Chilled Vater (P50) System P&ID's. This change- .
specifies pipe caps on all normally closed vent, drain, and instrumentation ~ lines. 1 Summary: I. No. This design change shows caps on all standard root valve' . configurations which are normally closed for vent, drain and v' instrumentation lines. This parallels the plant configuration. Therefore, based on the fact that overall system design.and function-has not been changed, the parameters upon which the accident analysis in the USAR was based, have not been affected.
-II. No. .The pipe caps are another means of isolation on a line which is ,
normally closed. There is no change to system function / operation.- Therefore, malfunctions of a different type vill not be created. III. No. Showing the caps on the P&ID's is an editorial function. System function / operation' remains the same. . Therefore, the margin of safety as specified in the Technical Specifications have not been' affected. h l l i, t
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~ . 1 SE No.r. 89-151 '
Source Document: NR NEDS 4163 ' Description'of Change
-This Safety Evaluation Analyzes the Nonconformance Report of the Division 1 Diesel Generator Field Ground Relay (64F) not being available to annunciate a field ground when the remote controls are isolated.
Summary-t I. No. At present the Generator Field Ground Relay (64F) vill not be ' available to annunciate the field ground when the remote controls are isolated. This relay is used,for alarm only purposes and does
, not have'any trip function. The diesel can operate indefinitely n
with a single ground existing on the field. The only conditions for which the diesel vould be operated locally (field ground alarm , inoperable) would' be testing or during a Control Room evacuation. Further, under these conditions no automatic starts due to undervoltage or LOCA signals are available. Operating with.an unknown ground would add a slight additional' commercial risk of generator damage. For a Control Room evacuation, the diesel vould be run even with a knovn ground if it was needed. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. The Field Ground Relay vill not cause or prevent the occurrence of'a field-ground. The Relay provides only an alarm function. The l non-operability of this' relay therefore, vill not create the , L possibility for an accident or malfunction which is outside of the USAR' evaluation.
.III. No. The Generator Field Ground Relay (64F) is not specifically discussed in the Technical Specifications. The Technical Specifications bases for electrical power sources are to ensure adequate power availability to supply safety-related equipment for safe shutdown and accident mitigation. The lack of an operable field ground relay vill not adversely affect these bases. Therefore, the margin of safety is not reduced.
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~SE No.:- 89-152 i
Source Document: TAF 80270 TDI DR/0R Report Rev. 1, item DR-03-02-650B-0 Description of-Change This evaluation is being made to justify the use of the presently installed "K2" field flashing contactors for the Division 162 Diesel
' Generators.
r Summary I. No. Inspections and evaluation of test data indicate that the presently installed K2 field _ flashing contactors are adequate to ensure proper-field flashing. Therefore, the probability of occurrence or the consequences of'an accident or malfunction of equipment important to safety previously evaluated in the USAR are not increased. II. No. Because there is no change to the plant,'the possibility for an l' accident or malfunction of a different type than any evaluated previously in the USAR is not created..
-III. No. There is no change to the plant. The margin of safety as defined in the basis of the Technical Specification is not reduced. l
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-SE No'.: 89-153
, Source Document DCN 2785
' Description of Change These drawing changes involve clarifying the USAR figures to properly depict the position of penetration test connection valves.
USAR Figure 9.3 Post Accident Sampling System i USAR Figure 9.2. Mixed Bed Demineralizer and Distribution System USAR Figure 9.4 Containment Vessel Chilled Vater System f USAR Figure 9.5 Fire Service Vater USAR Figure 6.2 Containment and Dryvell Isolation Summary I..No. These drawing changes reflect the same basic equipment configuration and safety classification for each of the penetration test ] connections valves.- The valves-vere previously designated or ! implied to be Normally Closed (NC) with the associated test j connection capped. The change to Locked Closed (LC) has no affect I on system operability and improves system integrity by strengthening . administrative controls. Therefore, these changes vill not increase } the possibility of the occurrence or consequences of an accident or ! malfunction of equipment previously evaluated in the USAR. ! II.'No. See item I above. l i III. No. These changes do not alter the function or operation of the systems ' in question. Administrative controls are strengthened. Therefore, the margin of safety in the bases of any Technician Specification is not reduced. l l i i
r i Attcchm:nt 3 PY-CE1/NRR-1141 L Page 232 of 285 SE No.: 89-154 Source Document: USAR CR 88-196 Description of Change This change clarifies the active / passive function of the containment and dryvell isolation dampers during Modes 1, 2, and 3. Summary 'l
-I. No. This change is administrative in nature.- It clarifies the Dryvell/ Containment isolation damper positions during various plant modes of-operation. This change does not alter damper positions, Therefore, the probability of occurrence or. consequences of an-accident or malfunction of equipment previously evaluated is not increased.
II. No.,-See item I above.
,a III. No.. This change provides clarificat' ion of the required damper positions, it does not change the positions. Therefore, the margins of safety
' defined in the Technical Specifications is not reduced. l ll 1
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F4 Attachinnt 3 .. PY-CEI/NRR-1141 L p Page 233 of 285 SE No.: 89-155 Source Documents USAR CR 89-094' , Description of' Change This change request' updates USAR Figure 11.5-1 (Sheet 5 of 12), Plant Radiation Monitoring, as a result of a-piping modification. j Summary L I.-No. This drawing change is editorial in nature.: It updates USAR Figure 11.5-1 to reflect current field conditions. The actual piping modification being depicted was previously evaluated as satisfactory in DCP 86-625. Therefore, the probability of occurrence or the-consequences of an accident or malfunction of equipment is not increased. ' e II. No. This drawing change reflects the current as-built condition of the plant. The actual design change was previously evaluated as satisfactory.- Therefore, the possibility for creating _an accident or malfunction different-than those evaluated has not increased. III. No. This drawing change is editorial only and reflects the current as-built condition of the-plant. Hence, the margins of safety described-in the Technical Specifications is not reduced. t i r
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? , [i I Attcch ;nt 3 PY-CEI/NRR-ll41 L , Page 234 of 285 SE No.: 89-157 , Source Document: DCP 88-0212, Rev. 0 Description of Change t This design change adds a LOCA~over-ride function to valve IP52-F646 to' . facilitate the restoration of Instrument Air (PS2) to the dryvell upon receipt of an isolation signal. (Electrical Evaluation). Summary
'I. No. This design change permits restoration of-instrument air to dryvell components following a LOCA. This modification does not alter the e
isolation function of the containment and the.dryvell. The electrical components-added by this change are seismically and environmentally qualified for this application (File SP568-000-00). The design is consistent with IEEE Standards 323 and 279.- Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not. increased.
.II. No. The nov control option affects a single dryvell isolation' valve.
Single failures of isolation valves are already addressed in the USAR Chapter 6. The override of the LOCA isolation signal can only occur after the valve has isolated during a LOCA. Hence, the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR is not created. III. No. This new control option does.not affect containment integrity or isolation as described in the Technical-Specification bases. Therefore, the margin of safety as defined in the bases for any .i Technical Specification is not reduced.
Attach::nt 3' PY-CEI/NRR-1141 L p Page 235 of 285 SE No.: 89-158-Source Document: DCP 88-0212, Rev. 0
- Description of Change This design change adds a LOCA over-ride function to value 1PS2-F646 to-facilitate the restoration of Instrument Air (PS2) to the Dryvell upon receipt of an isolation. signal. (Mechanical Evaluation)
Summary
-1. No. This design change alters the control circuitry of a currently installed dryvell. isolation valve. The modification does not alter- !
the design or the function of the valve. Therefore, the probability of the occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated is not increased. j II. No.~ This design change only affects the control circuitry.of a single isolation valve.- Single isolation valve failures have already been analyzed (Reference USAR Chapter 6). Hence, the possibility of an accident or malfunction of a different type than 'previously evalt;ated is not created. j i III. No. This new control option does not affect containment integrity or 3 isolution as described in-the Technical Specification. Therefore, ! the margin of safety as defined in the bases for any Technical } Specification is not reduced, i a 1 i j
w i Atttchzent 3 PY-CEI/NRR-1141 L-Page 236 of 285 SE No.: 89L159
- Source Document: FTI B01, Rev.E1, TCN 1
' Description of Change The control rod scram time testing procedure is being changed to provide an alternate method. For peripheral corner rods, that do not have two adjacent rods in either direction, it allows that only one adjacent rod next to the subject rod must be inserted for the scram test. It applies to the 20 corner rods out of.a total 177 control rods in the core.
Summary
..I. No. A study has been performed to ensure that this alternate method for corner rods does not violate'the control Rod Drop Accident (CRDA) analysis. The probability of occurrence on the consequences of an accident or malfunction of equipment previously evaluated in the
>- -USAR is not increased. II. No. This method is bounded by the CRDA analysis. 'The possibility for an accident or malfunction of a different type than any previously evaluated-in the USAR .is not increased. "3 i III. No. The bypassing of RPCS inhibits per the provisions of this instructJon is performed at less than or equal to 20% of RATED THERHAL F.,WER. 'i The bases of the Minimum Critical Pover Ratio (MCPR) Technical Specifications state that there is considerable margin to the MCPR operating limit of all designated rod patterns at all conditions less than or equal to 25% of RATED THERHAL POWER. ! Therefore, the margin to the MCPR safety limit is not reduced. I i 4 s i
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SE No.: 89-160- , L - Source Document DCN 2806, Rev. 0-Description of Change "
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This drawing change involves correcting the relief valve setpoints of the Steam Seal. System, to agree with General Electric's (GE) pipe drawings. L. Summary.
- I . No .- This drawing change simply' corrects relief valve setpoints to agree with.the vendor drawing and the field installation. This drawing change vill'not increase the probability of occurrence or the .
consequences of an accident or malfunction of equipment important to ' safety.
'II. No.- This drawing change vill'not. change the original function of the -
Steam Seal System.. Therefore,_the chance of an accident or ' malfunction of a different.- type than any evaluated previously in the USAR does not exist. ' III. No.- This drawing change vill not affect the operability or availability of the Steam Seal System or any system described in the Technical Specifications. The margin of safety as defined in the bases for any Technical Specifications vill not be reduced. a l
Attcchr;nt 3 PY-CEI/NRR-1141 h Page 238 of 285 SE No.: 89-161 Source Documents SCR's 1-89-1386, 1-89-1387 Description of Change This evaluation analyzes setpoint changes which increase the turbine feedpump high speed stop settings for 1N27-J003A/B from 5,450 rpm to 5,610 rpm. Summary
- 1. No. USAR Section 15.1.2 discusses feedvater controller failure-maximum demand and the resultant maximum feedvater runout flow to the reactor _ vessel. The probability of this malfunction is not i
, increased by this_setpoint change, since the feedvater controller is I a separate device from the high speed stop adjustment and vill not be touched or affected by this adjustment. Although the maximum feedvater flow, with this new setpoint, which-vould occur during -
this transient is greater (136.9%) than that used in the analysis in Section 15.1.2 (130%), the consequences of the transient will remain within the results and bounds of the original safety analysis. Further, a feedvater runout flow of 143% vas used in the transient analyses performed by GE for the Reload Submittal to the NRC. These analyses vere transmitted to the NRC under PY-CEI/NRR-0935L and approved by the NRC under Docket 50-440. Therefore, the probability of occurrence or the consequences of an accident or malfunction of
. equipment previously evaluated in the USAR is not increased, t
II. No. The setpoint changes do not affect the normal operation or function of the feed system, but just sets a new limit on the maximum speed
- of the turbine feed pumps. No hardware changes have been made.
He'ce, n the' possibility of a new or different accidents or L malfunctions _than previously evaluated is not created-III. No. See Item I above concerning the reload submittal. t M I
H p L Attechn:nt 3 4 PY-CEI/NRR-ll41 L Page,239 of.285 SE No'.: 89-162 : Source Document: SXI-041, Rev. O Description of Change Performance of this special test is the last in a series which evaluated
, anomalous reactor level instrument behavior during Reactor Core ~ Injection Cooling (RCIC) flov to the reactor pressure vessel. This testing is-confirmatory to ensure that the permanent modifications to-the reactor ,
level instrumentation is sufficient to preclude any anomalous behavior. , Summary I. No. This special test instruction is intended to provide data to verify > that permanent modifications on the level instrument reference legs ( successfully eliminates reactor level indication anomalies encountered during RCIC injections. This testing is enveloped by the system and plant requirements described in the original startup test program contained in USAR Chapter 14. The operation of the RCIC System for test is in accordance with approved plant operating procedures. Operation of level instrumentation during performance of the test vill be in accordance with Plant Technical Specifications. Therefore, the possibility of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in
.the USAR has not increased.
II. No.- See Item I above.
- III. No. This' instruction vill not affect the operability or availability of any system described in the Technical Specifications. Hence, the margin of safety for any Technical Specification is not reduced.
Attachm:nt 3 PY-CEI/NRR-Il41 L Page 240 of 285 SE No.: 89-163 Source Document: FCR 12861 Description-of Changes-LThis Field Change Request installs a freeze seal to perform maintenance on Reactor Vater Cleanup System valve IG33-F102. f Summary _; I . No .' This freeze seal is being used to perform maintenance on valve IG33-F102. This technique vill not' permanently alter the piping system. Valve 1G33-F102 has been previously disassembled without n' freeze seal. Vith the upstream valves closed, leakage past those valves is 5 GPH. This value is within plant make-up capabilities. The freeze seal vill be installed with the plant in Mode 4 (Low Reactor Vater Temperature / Pressure). If the line at the freeze seal-vould severe, leakage would be limited to the leakage past the closed upstream valves. Further, two ECCS systems are - required by Technical Specifications to be operable in this operational mode. Therefore, the possibility of an occurrence or consequences.of an accident previously evaluated in the USAR is not increased. II. No. See Item I above. ;
'III. No. .This; freeze seal and its potential failure has been analyzed and it '
vill not affect reactor vessel vater level. Therefore, the margin of safety as defined in the bases of Technical Specifications is not reduced.
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4 PY-CEI/NRR-1141 L EC Page 241 of 285 w SE Nod 89-164 Source Document: DCP 89-0144, REV. O Description of Change This design change vill provide additional air circulation in the upper
.dryvell region (Elevation 652'-0") by adding two take-offs to each of the two existing 13-inch dryvell cooling system round ducts.
Summary.
-I. No. The Dryvell Cooling System is seismically supported but is not '
required for safe shutdown of the plant. This modification vill L. take airflow from the dryvell dome area and re-distribute it to 1below the reactor vessel bulkhead plate and around the refueling bellows. This~is to add additional air circulation'and ultimately. some additional heat removal. Calculations using the original design-basis and using the operational data measured during the 1st cycle showed that reducing the flow to upperdome by 500 cfm vill not adversely affect the dome temperatures. Thus, the Technical
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Specification limit of 145'F for normal operation, USAR limit of 150'F (localized)'and the 330'F posteLOCA temperatures are not affected by this change. Therefore, this. modification does not increase the probability of an accident or a malfunction. II. No. The Dryvell Cooling System is not required to safely shutdown the plant and this modification does not change the overall cooling capability of the system. Therefore, the possibility for an accident different than that previously evaluated in the USAR is not created.
. III. No. This modification only re-distributes air flow within the upper.
dryvell and vill not affect Technical Specification 3.6.2.6 maximum dryvell average temperature of 145'F. Thus the margin of safety is not reduced.
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-SE No.:-- 89-165
. Source Documents. DCP 89-0144, Rev. O Description of Change i
.This evaluation analyzes the increase in Environmental Zone DV-4
,7 temperatures caused by dryvell system modifications. Summary I. No. Auditable File Package (AFP) E-627-000-01 envelopes the nev temperatures.. Hence, equipment qualification is not affected. Therefore, the probability.of occurrence or consequences of an accident or malfunction of equipment is not increased. II. No. 1There is no affect upon equipment qualification by changing the anticipated normal temperatures in Environmental Zone DV-4.- Hence, the-possibility of an accident or malfunction of a'different type is ' ' not created. III. No. ' Technical Specification margins are not.affected by this. change since the AFP envelopes the new temperatures. I
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Attachm:nt 3 PY-CEI/NRR-1141 L g Page 243 of 285 i SE NO. . 89-166 Source Document: PAP-0103, Rev. 4 Description of Change: This evaluation analyzes changes made to PAP-0103, Plant Operations , Review Committee. The changes make the Plant Operations Review Committee (PORC) membership and activities consistent with Amendment 22' ' of the Facility Operating License. Summary-I . No. - This is an administration procedure that describes the membership - and activities of PORC. It is consistent with Amendment 22 of the Operuting License. .Therefore, it vill not increase the possibility of occurrenc~e or consequences of an accident or malfunction i previously evaluated. . II. No. .This change is an administrative change, therefore the possibility-of an accident or malfunction of a different type than any evaluated previously in the USAR is not created. *
~ III. No. 'This change in PORC membership and activities is consistent with .
Amendment 22 of the Facility Operating L-icense. Hence, the margin ' of safety as described in Technical Specifications is not reduced. l l
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. c -PY-CEI/NRR-1141 L Page 244 of=285 JSE No.: 89-167 Source Document: ' USAR CR 89-117 Description of Change This change updates licensing commitments 7, 14, and 16 in bSAR Appendix 1B based on conversations with NRC and our follow-up letter
-dated March 3, 1989.
Summary I.'No. These changes all involve delays in NRC review.of various documents-and testing methodologies. NRC has authorized interim operations . With present configurations by-concluding that.the present .
. configurations are. safe 1This further delay has been justified in the same manner by NRC. : . .II. No. These delays have been determined to be acceptable by NRC. !
Therefore, they vill not increase the probability of an accident or malfunction of a.different type than any. evaluated previously in the USAR. III. No. No changes to design an operation vill occur as a result of these delays. Therefore, the margins of safety as defined in the bases of the Technical Specifications are not affected.
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.SE No.: 89-168 Source Document: DCP 88-0339, Rev. 4 Description of Change This revision of the design change alters the design to have the original loop seal level svitches (ITT Barton) provide the high alarm functions for the holdup line, the prefilter, and the cooling condenser moisture separator loop seals of the Offgas (N64) System. The lov level alarm and the valve. control function signals vill remain as designed as indicated in'the previous revisions of this design change.
l Summary j I
- I.- No. 'This design revision uses the existing ITT-Barton level
~ instrumentation to provide local indication and a high level alarm
" in the Control Room. The new instrumentation (Drexelbrook) vill provide control to the automatic loop seal shutoff valve and a lov level alarm in the Control Room. .This revision to the design is consistent with the original design requirements of the N64 syste' The system ~ analysis shows that the failure of the offgas-loop seals vill not compromise any safety-related systems or. prevent safe 4 shutdown. Therefore, the probability of occurrence or consequences !
of an accident or malfunction of equipment important to safety , previously evaluated in the USAR is not increased. ;
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II. No. The possibility for an accident or malfunction different than those. 1 previously evaluated has not been increased because the same design 'l' and installation parameters as the original design vere taken into account in the evaluation of this design revision. -l lII. No. This design revision vill not affect the margin of safety defined in i the bases of the Technical Specifications (Section 3/4.11.24 Gaseous { Radvaste Treatment (Offgas) since the system function has not been 4 changed. l 4 i l
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g k Attechm:nt 3 PY-CEI/NRR-1141 L Page 246 of 285 SE'NO.: 89-169. Source Documents. USAR CR 87-155 Description of Change: This change corrects inconsistencies in USAR Section 6.5.3.2.3 concerning the malfunction of components in the Annulus Exhaust Gas s Treatment (AEGTS) system. Summary-I. No. AEGTS has two 100% capacity, completely independent and redundant trains. If a component in the operating train malfunctions, the standby train is automatically started and placed in service. The standby train vill operate satisfactorily and4 perform its intended function irregardless of the continued operation of the malfunction i train. Thus failure of an active or passive component of one train and the' continued operation of that train vill have no impact upon overall system operability. Therefore,.the probability of' occurrence or the consequences of the accident or malfunction of equipment important to safety previously evaluated in the USAR is ! not. increased. 1 II. No. The-failure of an active or passive component of one train and ; consequently continued operation with that failure vill have no affect on the operation of the other train. Thus AEGTS System: LJ operability is ensured. Therefore, the possibility for an accident I or malfunction of a different type than evaluated previously in this 1 USAR has not been created. , III. No. There is'no change to the operation of the AEGTS System as described l' in Technical Specifications. The margin of safety as defined in the bases for Technical Specifications is not reduced. I l 4 l
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Page 247 of 285 SE No.: 89-170 replaces 87-539 Source Document: DCP 87-0039, Rev. O Description of Change p The Auxiliary Building flood detection level switches (IG61N515A/B) y located in the floor drain sump are highly inaccessible, present an ALARA concern during calibration, and are prone to damage due to the enclosed sump arrangement. This design change replaces the existing Magnetrol level switches with FCI switches. (I&C Hechanical Evaluation)
~ Summary I. No. 1G61N515A/B flood-level detection switches provide ;
a nonsafety-related alarm function only. Safety-related switches were provided in the initial design only to provide a more reliable switch compared to a commercial grade switch. All other components , in the alarm circuit are nonsafety-related. The installation of the new FCI switches will enhance the reliability of the installation due to the fact the switches are now accessible, have no moving parts, do not present an ALARA concern during calibration and are not prone to damage due to the sump environment. The level switches are not connected or tied to any equipment required for safe shutdown of the plant. The new level switches vill perform the same function as the original level switches and are considered a direct replacement as far as function. . Therefore, the probability of occurrence of the censequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR in NOT increased. II. No. The switch replacements vill not increase the possibility for an accident or malfunction of a different type than any evaluation previously in the USAR. The Icvel switches are not connected or tied to any equipment required for safe shutdown of the plant. III. No. The flood level detection portion of the Liquid Radvaste Sumps (G61) System is not described in the Technical Specification and no-reliance is placed upon G61 for plant operability. (Reference Technical Specifications 3/4.3, 3/4.4, 3/4.5, 3/4/.7, 3/4.8 and the bases 3/4 1-12.) The margin of safety as defined in the basis of the Technical Specification is not reduced.
__ .+- , Attachm:nt 3 PY-CEI/NRR-1141 b Page 248 of 285 SE No.: 89-172 Source Document: SOI N64/62, Rev. O, TCN-4 Description Change
'This temporary change to System Operating Instruction N64/62 gives operating instructions for the operation of the charcoal adsorbers at ambient temperature and operation of the charcoal adsorbers with one-train in service and one train out of service.
Summary t I. No. USAR section 11.3.2.1.6.1 allows limited operations of the charcoal adsorbers above 40' F. It specifies that this may occur due to maintenance activities on the refrigeration system. The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety is not increased by operation of the charcoal adsorber in a mode allowed during maintenance on the refrigeration system. II. No. See I above. III. No. The margin of safety of Teshnical Specification 3.11.1.2 is not reduced by operating the charcoal adsorbers at a higher operating temperature. Gaseous effluents vill be increased by operating at a higher temperature or by operating with a single train. By limiting condenser leakage and/or by limiting operating time at ambient temperature and by limiting single train operation the total annual dose to a member of the public can be kept within the requirements of Section II.A and IV.A of Appendix I, 10 CFR Part 50.
( t Attcchm:nt-3 PY-CEI/NRR-1141 L Page 249 of 285 ma
- SE'No.: 89-173
. Source Document: DCP 89-0168, Rev. O Description of Change This design change adds a capped sockolet (socket veld pipe adapter) to
.the Ofigas System, Charcoal Absorber (IN64-D0012A). inlet piping to allow for: piping internal inspection.
n
- Summary I. No. The. addition of a capped socknlet to 1N64-D0012A is_being performed ;
in accordance vith the system design. requirement in USAR section 11.3.2.2.1.6 for a 2-inch connection. Materials are in conformance h with USAR Section 11.3. Helium leak testing is not necessary I following modification because the socket velds are considered to be ' adequately assured of leak tightness by means of liquid penetrant testing.. Therefore, the probability of the occurrence or .( consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. The only-potential malfunction or accident which could be caused by > this change vould be a pipe break. This size break 2-inch is I bounded by the events analyzed in chapter 15.7 of the USAR. . III.:No.- Charcoal Absorber 1N64D0012A continues to meet all design and
-performance required as stated in Technical Specifications. The margin of safety as defined in Technical Specification'3/4.11.2, Caseous Effluents is not reduced.
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- I Attschr:nt 3 PY-CEI/NRR-1141 L Page 250 of-285
' SE NO. 89-174~
Source Documenti- SXI-0039 Description of Change-
-This instruction implements the Feedvater Pump Runout Capability Test which was performed on Reactor Feedvater Pump B~due to impeller [
replacement.
- Summary I
I . No .' USAR Section 15.1.2 and Appendix 15E discusses a feedvater controller failure ' maximum feed water runout-flow to the reactor vessel. This test verifies the capability of the feedpumps and, if , necessary, adjusts the high speed stops of the Reactor Feed Pump ' Turbine-(RFPT) to the level necessary to not exceed the maximum
.feedvater flow to the vessel. The maximum feedvater flow used by this-test is 143%. This is greater than_the 130% currently listed Ti in the USAR. This increased flov was used by GE in the Reload' ,
Submittal to the NRC that vas transmitted to the NRC via PY-CEI/NRR-0935L and approved by the NRC under Docket 50-440. Therefore, the probability of the occurrence or consequences of an accident'or malfunction of-equipment previously evaluated is not ^
. increased. :
i II. No. This test verifies the feed vater runout flov and makes the ' necessary_ adjustment to the RFPT high speed stops to not exceed the new maximum feedvater runout flow of 143% that the NRC has approved. III. No. 'All Technical Specification changes requiring the use of the 143% I feedvater runout flow have been approved by the NRC. These changes were evaluated to the same criteria as the original plant design. The results demonstrated that the consequences of the transients or accidents did not increase beyond those previously evaluated by the - NRC. !
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- PY-CEI/NRR-1141 L
{ Page 251 of 285 SE No.: 89-175 i Source Document: USAR CR 89-153 ' p
",' hscription of Change This change evaluates changing the minimum temperatures for Environmental Zones DV-3 and DV-5 to 80'F.
i [ Summary I. No. Per GE document 22A2741, Rev. 2, the inlet cooling air temperature to the vessel support skirt shall not exceed 120'F and shall nct be lover than 100'F. The response to FCR 10873 vhich evaluated the t
; impact on the reactor vessel skirt for the life of the plant at belov operating temperatures stated that the reactor vessel skirt could experience an operating temperature as lov as 80'F (at sensor 3"
1M13-R110) for the plant's life and without adversely affecting the vessel. To decrease the temperature to the Reactor vessel support skirt the ' Nuclear Closed Cooling (NCC) water flow has to be increased to the lover dryvell coolers. Increasing this cooling water flow vill not have an adverse affect on the NCC System since the contrcl valve has been operating only 40% open. Therefore, decreasirg the minimum temperatures for the under vessel area and the reactor vessel skirt area vill not increase the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated in the USAR. II. No.
. See Item I above.
III. No. This change vill not affect Technical Specification 3.6.2.6. Therefore, the margin of safety described in the Technical Specifications is not reduced. - L
h
' ',' Attcchment 3 PY-CEI/NRR-1141 L Page 252 of 285 SE No. 176 Source Document: DCN 2816, Rev. O Description of Change This drawing e.hange revises the Piping and Instrument Drawing (P & ID) for the Containment Integrated Leak Rate 7esting System to reflect the physical installation of flov element IE61-FE-N444.
Summary I. No. Flov element 1E61-FE-N444 is located in the nonsafety portion of the Containment Integrated Leak Rate Testing System. This portion _of the system is used for conducting the containment Integrated Leak Rate Test (ILRT) prior to startup and is no vay utilized for safe shutdown of the reactor nor is it part of the boundary required to establish containment integrity. Reference USAR Section 6.2.6.1 and Figure 6.2-65. The as-installed location of flow element vill have no effect on the performance of the Containnient Integrated Leak Rate Testing System. This is based on successful completion of prior ILRT's. Therefore the probability of occurrence or the consequences for an accident or malfunction of equipment important to safety previously evaluated in the USAR does not increased. II. No. See Item I above. III.- No. Flov element IE61-FE-N444 is located in the nonsafety portion of the Containment Integrated Leak Rate Testing System and is not required for plant operation, safe shutdown, nor is it part of the containment boundary. .The margin of safety as defined in the bases for Technical Specifications 3/4.6, containment Systems and 3/4.6.1.6, Crntainment Internal Pressure, is not reduced. i
Attachment 3 PY-CEI/NRR-1161 L [ Page 253 of 285 SE No.: 89-177 ! Source Document: PEI-B13, Rev. 1, TCN-10 f Description of Change ! This change incorporates new values for the Minimum Alternate Reactor " Pressure Vessel (RPV) Flooding Pressure (MARFP). Amendment 20 to the PNPP. Operating License revised the Linear Heat Generation Rate (LHGR) i limits as a result of the new fuel added during first refueling outage l for cycle 2. This resulted in a review of the calculations used to support numbers and graphs in the Plant Emergency Instructions (PEI's). The review of the calculations resulted in these changes to the MARFP pressures. The Minimum Alternate Flooding Pressure is defined as the lowest RPV > pressure for a given number of open Safety Relief Valves (SRV's) at which ; steam flow up through a completely uncovered core and out the SRV's can adequately cool the core by heat transfer to the steam. Once RPV f pressure drops below the MARFP, the rate of depressurization is small and . injection into the RPV must be re-established.in order to adequately cool the core and increase RPV vater level. I Summary , I. No. This change corrects the PEI's so that they are applicable to the new fuel installed for Cycle 2. The PEI steps involved are part of the plant recovery actions after an accident or malfunction has < i occurred. -They do not increase the probability of occurrence or the . consequences of an accident or malfunction evaluated in the USAR, but are discussed in the Perry Specific Technical Guideline.
- II. No. These changes involve plant recovery actions after an accident or malfunction has occurred. Changing the values of the'MARFP table in r a more conservative and correct direction vill in no way increase the possibility of an accident or' malfunction eva)uated in the USAR. ,
III. No. These changes involve the value of RPV pressures at which operator actions during accident conditions should occur. They are beyond the scope of the Technical Specification basis. These actions are consistent with the Emergency Procedure Guidelines. . f i t 9 I i 1 l
1 Attcchment 3 PY-CEI/NRR-1141 L Page 254 of 285 n
! SE No.s. 89-178-Source Documents .SXI-0038, Rev. 0 Description'of Change-
- p. ' "
3 (. Special Test Instruction, $XI-0038, is a temporary test procedure that modifies the operating line up of the Turbine Building and Heater Bay
;. . Ventilation Systems.to verify the building airflow directions. The test vill be performed'in Mode 4 (shutdown) or 5 (refuel). SXI 0038, Rev. 0-expires 3-3-90.
- Summary No. This procedure vill not alter the original function of the Turbine
[ 'I. Building and Heater Bay Ventilation systems. Further, this procedure does not modify the operation of any safety-related system. Therefore, the probability of occurrence or the consequences'of an accident or malfunction of equipment important to L safety previously. evaluated in the USAR has not. increased. II. No. See Item I above. III. No. Performance of this procedure vill not reduce the margin of safety as described in the bases for any Technical Specification. s I i i l' l , t l w~ .-
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Attcchment 3 PY-CEI/NRR-II41 L 7 Page 255 of 285 l SE No.: 89-179 Source Document: USAR CR 89-069 Description of Change This change replaces a vendor drawing in the USAR vith a P&ID. The existing USAR Figure 9.5-25 currently reflects the vendor's generic P&ID - . for the High Pressure Core Spray (HPCS) Diesel Generator (DG) Lube Oil ' system. This. drawing has been revised and reformatted to reflect f as-built system details. ; Summary I. No.- These changes are limited to editorial and minor configuration [ changes to reflect the as-built plant. This system still conforms
- J to the original design codes and standards. Therefore, the ,
probability of occurrence or.the consequences.of an accident or l malfunction of equipment important to safety previously evaluated in the USAR-has not been increased. ! II. No. These changes merely reflect the as-built plant which has been -! previously evaluated. There is no increase in the possibility for i an accident or malfunction of a different type than any evaluated previously. ,
. 111. No. These changes do not affect Technical Specificationo nor do they 3 reduce the margin of safety as defined 'a the bases for any 1
. Technical Specification, j f
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"" Attcchment 3 -
I PY-CEI/NRR-1341 L , Page 256 of 285 ; i SE No.a. 89-180 :i Source Document: USAR_CR 89-093 l Description of Change p This evaluation analyses an editorial change to Chapter IA'and an ; E addition to Chapter 6.2 Section 5.2 tegarding the reference associated , [. vith the hydrogen analyzers. ; Y Summary l L[- ' I .- No. The changes to Appendix 1A are editorial and simply change the tense : of verbs since the action to install hydrogen analyzers required by- j NUREG-0737 has been taken. Hydrogen analyzers are started upon entry into PEI-B13 to support mitigation of hydrogen producing events by providing an entry t rondition for PEI-M51/56 on hydrogen concentration. This action is 6
'taken to satisfy requirements in 10 CPR 50.44. The statement added t to Chapter 6.2 repeats the commitment to requirements imposed on the ,
hydrogen analyzers NUREG-0737 established-in Appendix 1A of the USAR._ These requirements do not impact the operability _of the , analyzers as defined in Chapter 6.2 and do not affect any of the , 4 actions assumed in the DBA. Therefore this change does not affect i the probability of occurrence or_ consequences of any accident 1 previously evaluated. t II. No. ~See Item I above. , L III. No.- This change does not impact the basis for any Technical Specification. l ! h a t t t I ( i 6 i r
r V,' .
/ AttOchment 3 !
PY-CEI/NRR-1141 L Page 257 of 285 l SE No.: 89-181 i p Source Document: USAR CR 89-151 ?
.. Description of Change < 5 I
This evaluation analyzes the change in testing the Main Steam Bypass ! Valves monthly rather than veekly. ! l Summary [ s I. No. In SSCR 10, NRC stated that cycling these valves monthly was { 3,i consistent with the valve manufacturer's recommendations and with ' current industry practice and therefore, is acceptable. Testing monthly is adequate'to identify potential problem. The probability + of occurrence or consequences of an accident or malfunction of equipment previously evaluated is not increased. . II. No. These valve have always'been tested monthly. Since there is no -' change to the frequency,-the possibility for an accident or ; malfunction of a different type is not created. t III. No. This. change is consistent with the Technical Specifications require ; testing every thirty-one days. ' i
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e [ , Attochment 3 + PY-CEI/NRR-1141 L' J Page 258 of 285 i SE No.:- 89-182 :
~ Source Documents- USAR CR 89-115 !
i Description of: Change ! p- .This change corrects USAR Figure 2.3-9, offsite and onsite maximum vind g directional persistence roses. u : [t Summary h I.'No. The change.is editorial in nature. It corrects the 60-meter NV
-direction from 37 to-31. This change does not affect any Design !
3 Base Meteorological Calculation.- A decrease in vind direction - persistence from a given sector vill not affect plant equipment ; important to safety.- The probability of the. occurrence or ; consequences of an accident or malfunction of equipment previously r ['u -evaluated is not increased. ; II. No. See Item I above. I This change-to the maximum vind direction persistence rose does not ! III.' No. (. -effect-any Technical Specifications margin of safety calculation. L s F r* I D E 4 I i b i
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-Atttch:cnt 3 PY-CEI/NRR-1141 L j
Page 259 of~285 , i SE No.: '89-183 l Source Document USAR CR 89-114 } Description of Change j
, This change provides corrections to the Seven-Site-Year data reported in :
f the.1983 through 1987 annual reports and the USAR. Data presented in the ! original tables vere generated from the computer code VETTEMP vhich has a maximum data period restriction of' ten calendar years. This code, however, was applied to the Seven-Site-Year data, which spans eleven l
. calendar years,-(1972-1982). As n. consequence, the resultant output '
k
. contained erroneous data for maximum and minimum values (both maans and extremes) which vere incorporated into subsequent meteorological reports. i Summary l P
- 1. No. The meteorological values evaluated vill not affect previously performed design basis calculations. Therefore, the probability of i occurrence or the consequences of an accident or malfunction of '
equipment previously evaluated is not increased. II. No. See Item I above, f III. No. The minimum and maximum temperature valves previously evaluated have not changed.- Hence, no Technical Specification margin of safety has ' been reduced. i r a
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i t E l', 5 s k iT l < , - - <
7 p Attcchment 3 [ PY-CEI/NRR-1141 L , i Page 260 of 285 SE No.:- 89-185-Source Documents PAP-0506, Rev. 1 Description of Change This safety evaluation analyzes the alteration of the OD-1 program on the Process Computer arrays to accept substitute data from symmetric Transversing Incore Probe (TIP) locations. L Altering the OD-1 program vill permit recalibration of the Local Power Range Monitors (LPRM) with a reduced number of TIP machines. A Technical Specification amendment has been issued (Amendment 25) that permits the use of functioning channels of the TIP System to provide necessary data for LPRM recalibration when up to ten of the TIP measurement locations are inaccessible or inoperable. Summary I. No. The core monitoring methodology is based upon symmetry of rod patterns and fuel loading. Rod pattern is sequence A. : t This change does not change the fundamental process involved in calibrating neutron instrumentation (LPRM's). The use of symmetric TIP detectors to provide substitute data for inaccessible TIP channels does not compromise the ability of the process computer to accurately represent the spatial gamma flux distribution of the reactor enre. . E Purther, the basic method used to calculate power and exposure ' distributions and fuel thermal limits has not been altered. The existing method for calculating core power and exposure distributions and fuel thermal limits includes provisions for monitoring the gamma flux distribution using core symmetry. .. The calibration of LPRM's using symmetric flux distributions result in LPRM data being within ,he normal unew tainty expected for calibrations not using this substitute data (Normal TIP System Operation). Consequently, this vill not adversely affect core , thermal limit calculations. This change does not alter the function, performance or operation of , l any safety system or safety-related equipment. The limitation on l' total TIP uncertainty ensures that the readings from symmetric TIP channels are equivalent. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated is not increased. 1 l
(j Attcchment 3 ' PY-CEI/NRR-1141 L Page 261 of 285 l SE No.: 89-185 (Continued) ; II.-No. The substitution of data into inaccessible TIP channels has no ,
'effect on any accident initiator, therefore this proposed change !
L .does'not create the possibility of a new or different kind of * { -accident:from any accident previously evaluated. *
, III. No. :This change does not involve a reduction in a margin of safety since l u
L the implementation is restricted by the Limiting Condition for. . Operation (LCO) to sequence A control rod patterns and the total TIP l uncertainty has been demonstrated earlier in the cycle to be within j the value assumed in the General Electric Reload Licensing Topical : Report - GESTAR II. ' L i F I e t b 1 i
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g - h - I Attcchment 3 PY-CEI/NRR-1141 L ' 4 Page 262 of 285
- SE No.: 89-186- t
. Source Document: LL&JED 1-89-183 -
' Description of Change This change evaluates disabling. Loose Parts Monitoring System (LTd3) l' channel-six to prevent spurious initiation of the loose parts locator and i running the locator out of paper. .
Summary . j' I. No. The LPMS has ten loose parts channels.vhich 1:tisfy Technical L Specification requirements plus two additional non-technical ' L specification channels which are mounted on the reactor : L recirculation pumps (Reference USAR Section 4.4.6.1.2). This ! h, ' modification disables one of the non-technical specification > L channels. This change vill not alter the function of the LPMS.- a 1
-Therefore, the probability of occurrence or the consequences of an !
-accident or malfunction of equipment important to safety previously
, evaluated has not increased. II. No. As described,in item I above, the system can still perform its design function. Hence, tl'e possibility for an accident or
- malfunction of a different type than evaluated is not created.
III. No. This channel is not a Technical Specification channel, therefore,- ! the margin of safety is not reduced, i i s h v J P f - .
., ,r-
n Attcchment 3 FY-CE1/NRR-1141 L i Fage 263 of 285 SE No.: 89-187 Source Document: USAR CR 89-160 h Description of Change This change removes the Ohio Fire Academy training requirements for the Fire Brigade as found in USAR 9A.5 Summary I. No. The onsite Fire Brigade training in safe system shutdgun and fire fighting, meets or exceeds the offsite training provided by the Ohio Fire Academy. 10 CFR $0 Appendix R does not require offsite training. The deletion of offsite training would not result in an accident or malfunction of equipment. II. No. Offsite fire fighting training at the Ohio Fire Academy is not site-specific. Therefore, tnat training does not have an accident or malfunction implication. III. No. The training leve1~provided for the Fire Brigade vill be unchanged. Therefore, the Fire Brigade training described in Technical Specification is not reduced. t i V i-
I Attachmant 3 I PY-CEI/NRR-1141 L Page 264 of ??5 SE No: 89-189 Source Document: NR PPDS 4239, Rev. 1 Description of Change This evaluation analyzes the use-es-is disposition for the Reactor Core Injection Cooling (RCIC) valve IE51-F0514 failed closed position. Vith 1E51-F0514 failed closed, level switch IE51-N010 is rendered inoperable. The purpose of this level switch is to automatically ensure that the RCIC steam supply line remains free of condensate via the drain pot. This same design function can be accomplished manually by proper alignment of the valves in the drain pot to the condenser. Summa _ry
- 1. No. Manually keeping the RCIC steam supply line free of condensate vill ensure that RCIC continues to meet its design function as described in the USAR and Technical Specifications. RCIC operation under initiation conditions vill not be compromised as drain pot drain valves 1E51-F025/1E51-F026 vill receive isolation signals on the RCIC initiation. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR vill not increase.
II. No. The alternate steam line condensate drain configuration does not alter the operation of the RCIC system. All components of the drain line are operated within design parameters. Therefore, no new failure mode is created. III. No. The margin of safety as defined in Technical Specification 3/4.7.3 is not reduced. RCIC continues to meet all design eid performance requirements as stated in Technical Specifications. l ii
e ,; , - 1 4
,, e Attcchment 3 "N'i' PY-CEI/NRR-ll41 L l
. Page 265 of 285.
- SE No.:' 89-190.
. Source Documents. Delaval Vendor Manucl-Revision Notice, MRN 3 1 k . . . Description of Change 1 bi
.This esvaluation documents exceptions taken to the maintenance ;
recommendations made by Soutt. vest Research Institute (SVRI) in their ' ,. final report "A Design Review of Delaval DSRV-16 Engines at the Perry : !' Nuclear Power Plant,"' dated October 14, 1985. ' Summary g < I. No. .These changes do not affect the reliability, redundancy, or function j F of-the diesel generators and therefore, does not increase-the j M , probability of occurrence or the consequences of an accident - previously evaluated. ' F . 1 II. No. .No changes to the plant are being made; therefore, the possibility for an accident or malfunction of a different type than any: U evaluated previously is not increased. _ III. No. 'The reliability of the orisite power supply is not being changed; therefore, margin of safaty in the bases of the Technical Specification 3/4-8 is not being changed. ( s I P 1 ( 5 6
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L. Attcchment 3 PY-CEI/NRR-1141 L Page 266 of 285 , SE No.: 89-191 Source Document USAR CR 89-176
-PAP-0507, Rev. 8. TCN 4 !
[ F Description of Change This evaluation analyzes changes made to USAR Sections 2.1.2, 6.4.1, ,!
~13.3.1, and 13.5.2, and Table 13.5-1. The change deleted _ operations i manual volume designators from the text and updated the table to reflect i ,e revised procedures preparation responsibilities. !
Summary $ ie ~! e I. No. The_ change is administrative in nature. Surveillance instruction ! preparation vill continue to be the responsibility of qualified L[ groups. 'Therefore, the probability of occurrence or consequences of- ; an accident or malfunction of equipment previously evaluated is not 1 increased. , The administrative nature of this change can not cause a different
'II. No.
type of accident or malfunction. l III. No. This is an administrative change only. Therefore, it vill not reduce the margin of safety as defined in the basis for any Technical Specifications.
- l 4
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AttOchacnt 3 PY-CEI/NRR-1141 L Page 267 of 285 SE No.: 89-193 Source Document: USAR CR 89-159 Description of Change This change modifies USAR drawings 9.5-1 and 9.5-2 by adding augmented quality flags'and revising HPL numbers. Summary I. No. This change only corrects MPL numbers and scoping flags. These items do not affect the function of any system or modify the scope of the Augmented Quality Pro 5 ram as described in the USAR section 9A.S. The probability of the occurrence or consequences of an accident or malfunction equipment previously evaluated has not incorporated. , II. No. This change.does not affect any item in the Fire Protection Program. Therefore,'the possibility of an accident or malfunction of a different type than evaluated in the USAR is not created. III. No. This change has no impact on the administrative aspects of the Fire Protection Program and does not affect alternate shutdown capability. . Therefore, the portions of the Technical Specification applicable to fire protection vill not be affected. f
e t Attachment 3
. PY-CEI/NRR-1141 L Page'268 of 285 i
- SE No: 89-194 j
Source Document: DCN 2852 Description of Change 7 This drawing change updates the one-line diagram, D-206-051, by adding I drawing references. i
.)
. - Summary I. No. This drawing change updates the'dr'ving a reference section only. It j does not increase the probability of.an accident'or malfunction of equipment.
j l II. No. This drawing change updates the drawing reference section only.- No possibility.of different. type of accident or malfunction is created j t
'by.this change. .
III. No. The margin of safety as defined in'the bases for any Technical . ! Specification is not reduced by this drawing change since it revises the drawing reference only.
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Attcchment 3
- 9. PY-CEI/NRR-1141 L Page 269 of 285 SE No 89-195 Source Document: OM12E Process Control Program, Rev, 4 TCN-1 Description of Change This change allows radioactive vaste to be processed in the truck bay of Radvaste Building when it is determined prudent. Normal processing of radioactive vaste is performed in the process pit of the solid radvaste area.
Summary I. No. The interface between the Perry Plant and the vendor system have not changed. However, periodically the processing of vaste vill be performed in the truck bay. This change does not alter Perry's ability to meet those critical items associated with the vendor / Perry interface detailed in SSCR Section 11.4.
- a. Means to contain spills or overflov Periodically vaste processing vill be performed in the truck bay which means the hoses, fill head, and the container vill also be located there. This area is surrounded by concrete valls on three sides that vill contain any postulated spill or overflow. The fourth side is the truck bay access door which vill be keylocked shut and have temporary curbing placed in front of it to stop any potential spills. Also any hosing or connections outside the fill aisle vill be placed in hose bags to stop any potential leaks or hose breaks. A floor drain, routed to the liquid radvaste system provides adequate drainage to handle any postulated hose break or leak. All floors and valls in the areas are coated to ensure decontamination.
Finally all processing vill be performed in NRC approved casks thus eliminating any concerns with liquid tank failure analysis.
- b. Liquid tank failure analysis implication:
All processing vill be performed in an NRC approved cask therefore the potential for a massive tank spill is eliminated.
- c. Radiological and effluent monitoring:
No changes are necessary in the previous response because the truck bay utilizes the same ventilation system as the fill aisle area. Ventilation from this area is routed through HEPA filters and charcoal beds prior to release to the environment via the Unit I vent. Radiological monitoring is provided for Regulatory Guide 1.21 gaseous discharges from the liners are processed through the vendor's offgas blover system as detailed before.
i L Attechm:nt 3 PY-CEl/NRR-ll41 L Page 270 of 285 SE Non- 89-195 (Continued)
- d. ALARA considerations:
All process and vendor related functions are still operated from a remote control panel. These panels are located behind a 2 foot thick concrete vall. Also the liner is located inside a cask with 1.5 inch thick lead filled stainless steel required to reduce doses belov DOT guidelines of 200 mr on the exterior of the package. Closed circuit televiolon is provided for remote viewing of the processing area. By employing the remote cameras, cask, and Health Physics procedures, the time spent in this area is minimized. No additional source terms are introduced to the radvaste system due to this change. Consequently, the estimated offsite doses as presented in Section 5.2.4 of the PNPP Environmental report vill remain the same and thus vill remain the same and thus vill not exceed 10 CFR 50 Appendix 1 design objective. No new possibility exists. The vorst case possibility of a tank failure outside the radiological control boundary was previously evaluated in USAR Section 15.7.2 and 15.7.3. Therefore the probability of occurrence or the consequences from an accident or malfunction of equipment to safety previously evaluated in the USAR does not exist.
- 11. No. See Item I above.
Ill, No. This change has no impact on the actual processing capabilities detailed in the Process Control Program. Therefore, Technical Specification 4.11.3 Solid Radvaste Treatment is not impacted.
, . -> Attcchment 3 ;
PY-CEI/NRR-1141 L l i- Page 271 of 285 l l SE No: 89-196 l Source Document: DCP 88-0163 Rev. O ; Description of Change
! This design change vill provide installation of sliding link terminal blocks in Control Room panels 1H13-P618 and 1H13-P629 for the Residual i Heat Removal (RHR) System to improve testing perforrance and personnel
-safety.
= Summary U I. No. During. Surveillance Testing Program (SVI's), leads were lifted n'
either at the relay or at the terminal strip in order to isolate the circuit under test. .This presented a hazard to the test technician and equipment in the Control Room panels. The. installation of sliding link terminal blocks to eliminate lifting leads from relays or terminal strips does not alter the function and operability of the RHR system. This design vill provide for a safer and-more reliable means of performing periodic tests as addressed in the SVI's and delineated per Regulation Guide (RG) 1.118, Periodic Testing of Electrical Power and Protection Systems. Implementation of the design change vill not compromise or lessen the degree of conformance to R.G. 1.118 as defined in the USAR. Therefore, the probability of the occurrence or consequences of an accident or malfunction of equipment previously evaluated is not increased. II.-No. The terminal block replacements vill not increase the possibility for an accident or malfunction of a different type than any evaluation previously in the USAR. c III. No. This change does not alter the function or operation of the RHR system. .Further, the change enhances the testability of the system I and components. Therefore, the margin of safety as defined in the l basis of the Technical Specification is not reduced. 1 L l i
[ Attcchment 3 PY-CEI/NRR-1141 h k Page 272 of 285 SE NO.: 89-197 Source Document: DCN 2234, Rev. O
! Description of Change:
!" This drawing change reflects the elimination of the internals from check valves IP57-F509A/B and IP57-F512A/B of the Safety-Related Instrument Air (P57) System. The design modification essentially converted this portion of the P57 system to a straight pipe, moving the safety class boundary from upstream of the small receiver tanks to just upstream of the newly installed larger air receiver tanks. These modifications were completed as a result of an earlier design change which added two large receiver tanks to the system. Summary I. No. The P57 System has been redesigned as a low pressure system. Consequently, two additional air receivers were added to hold s sufficient air volume. The size of these receivers is more than adequate to meet design requirements for system air volume. Furthermore, during the performance of-special test SXI-002, Rev. I the ability of these air receivers to hold sufficient air alone was verified. While a high pressure system, the nonsafety and safety portions of the P57 System vere separated by 1P57-F512A/B. As a low pressure system, the boundary is moved to IP57-F555A/B vith the former high pressure receivers IP57-A0001A/B and A002A/B not being required due to the boundary change, 1P57-F509A/B and IP57-512A/B are no longer required, llence, the removal of their internals has no impact on operability of the P57 System. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated is not increased. II. No. This change effectively causes the system boundary to be moved to IP57-F555A/B. Air receiver tanks A001A/B and A002A/B are not needed to supply air for ADS operation. A003A/B are sufficient for this F purpose. The only change in system operation resulting from this action is that the compressor duty should be reduced f.e., fever starts and stops with longer run times. This should improve compressor reliability. This does not induce another accident scenario which is not analyzed.
-III. No. As described in Items I and II above, no margin of safety is reduced.
e y E Attechment 3 PY-CEI/NRR-1141 L Page 273 of 285 SE No: 89-198 Source Document: DCP 88-213 Description of Change i This design change reconfigures the two Technical Support Center (TSC) Uninterruptable Power Supplies (UPS) such that one of the UPS vill serve as a primary power source supplying all of the TSC loads, with the other UPS in standby, ready to pick up the load if system faults occur. (Fire Protection) L Summary I. No. Vith the exception of loss of automatic fire detection and suppression during a 3 to 4 hour period during installation, this changes o
- n. Does not interface with any existing fire protection feature.
- b. Does not affect the basis for Appendix R.
- c. Adds only an inconsequential quantity of combustibles to the !
plant area. During implementation of the change all automatic fire detection and-suppression systems may be unavailable. Should this be the case, compensatory measures vill be in effeet to provide assurance that any fire occurring during this 3 to 4 hour period vill be promptly detected and extinguished. , Accordingly it can be concluded that both the probability of occurrence or the consequence of a fire is not significantly increased during installation of this design change. ! II. No. Based upon the description in item I, the design change is not related to any known accident mode or failure mechanism for important plant safety features. Therefore the change does not create a new type of accident or malfunction. III. No. Based upon the description in item I, there is no relationship ' between the change and.the basis for any Technical Specification. Therefore, the change does not reduce any Technical Specification margin of safety.
i3":a ' Attcchment 3' !
, PY-CEI/NRR-1141 L l Page.274 of 285 {
, SE Nos. 89-199 ;
Source Document: USAR CR 89-161 '
. Description of Change
-l h
This evaluation analyzes the reorganization of the Nuclear Quality -l
,. Assurance Department (NOAD). These changes include the combining of l Procurement and Administration. Quality Section with the Maintenance and !
Modification Quality' Control Section. Also Operational Quality Section f is renamed Quality Assurance Section. l [ i Summary ' i h I. No. .This change is administrative in nature. NOAD activities are not '
- k. changed, only reassigned. Therefore, the probability of occurrence ;
;,7 or consequences of an. accidents or malfunction.of equipment j ! previously has not' increased. ;
f: [. II. No. As described in Item I above, this organizational change vill not ! [ . create a different type of accident or malfunction. ;
.III. No.- This change is administrative only, it vill not reduce the margin of safety as defined in the basis for any Technical Specifications, i 5
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.PY-CEI/NRR-ll41 L Page 275 of 285 SE No. . 89-200 bp Source Document USAR CR 89-144 L- Description of Change This change revises USAR Section 5.4.7.2 to make it optional, rather than mandatory, to re-route Residual '7at Removal (RHR) Heat Exchanger condensate to the Reactor Core Isolation Cooling (RCIC) System.
Summary
- 1. No. USAR Section 5.4.7.2 states that when in the Steam Condensing Mode, I' the RHR heat exchanger condensate is initially routed to the suppression pool. but once the condensate quality attains the appropriate level, it is re-routed to the RCIC System. This section conflicts with item II K.l.22 in USAR Appendix 1A, which states that-condensato can be sent to either location. The deletion of this requirement to re-route flow to RCIC System would eliminate the reliance on RCIC System operation for the Steam Condensing Mode of RHR. The safety function of the RHR System is not affected by this change. Therefore,_the probability of occurrence or the
. consequences of an accident or malfunction of equipment previously evaluated has not' increased.
II. No. Since the safety function of the RHR system is not effected. No new accidents or malfunctions are created by this change. III. No. This change does not affect any safety margins defined in the bases
, of' Technical Specifications.
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F Attcchment 3 PY-CEI/NRR-ll41 L
- Page 276 of 285 i-SE No.: 89-201 Source Document: USAR CR 89-110 L Description of Change l t
[ This change clarifies USAR Section 7.7.1.12 in that the Dryvell Vacuum ! Relief motor-operated isolation valves are automatically opened by a negative differential pressure signal and deletes the requirement to > manually override the LOCA isolation signal. Summary I. No. This change is an editorial change to the USAR. This change deletes an implied override of a containment isolation signal. No procedural direction is given to conduct this override. This override automatically occurs if needed due to a dryvell. negative t pressure signal. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment previously evaluated has not increased. II. No. This editorial change deletes an implied override of a containment isolation signal. This override automatically occurs if needed due i to dryvell negative pressure. No new accident or malfunction is created.
, III. No. This editorial change does not affect the margin of safety for any Technical Specification. l i
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77- o Attcch;;nt 3 PY-CEI/NRR-1141 L l Page 277 of 285 ! SE No.: 89-202 , Source-Document: USAR CR 89-102 ( [ < Description of' Change USAR Section 7.3.1.1.5 requires that Residual Heat Removal ! valves 1E12-F024A/B be throttled as necessary. IE12-F024A/B are gate ~ r valves having seal-in control' circuits in both the open and close .
~ directions and therefore are not designed for throttling. This change :
vill delete the reference to 1E12-F024A/B as throttle valves, and ; L instead, reference that 1E12-F048A/B and 1E12-F003A/B vill be throttled- ' as necessary. .i 5
; - Summary. !
E . . tH I. No. During accident. conditions IE12-F024A/B vill auto close regardless [ h of their positions. This is an editorial change for clarity. j II'. No. 'No new accidents or malfunctions are potentially created by this [ change since the safety function of these valves are not affected in any vay. !. III. No. This change does not conflict with any safety margins defined in.the bases of the Technical Specifications. . (. , i s L i e i t E
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, - 1 Attochment 3 cr PY-CEI/NRR-1141 L }! ,
Page 278 of 285-ll ' SE No. . 89-203 Source Documents .USAR_CR 89-061 Description of Change
- c1 This' change corrects and clarifles the USAR system description of the 4
' Nain~ Steam Isolation Valve Leakage Control System (MSIVLCS) operation.
- Summary
., I. No. This~ change simply clarifies conditions which require MSIVLCS ty operation, the change does not effect the probability or consequences of a LOCA.
II. No.-1The operation of MSIVLCS has not changed. Therefore, the l~ possibility for an accident or malfunctions of a different type than L evaluated is not created, c \ i
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W .' III. No. Technical specification margin of safety is not effected by this change.- j.. I ,. 'F [} ' 7 6, k __
_ ~ - Attcchm:nt 3 PY-CEI/NRR-ll41 L Page 279 of 285 SE No 89-2"4 Source Document: DCP 88-371, Rev. O Description of Change This design change allows for the replacement of the Reactor Vater Cleanup (RVCU) System multi-channel temperature transmitter 1G33-N003 with four single channel units for heat exchanger temperature monitoring. Summary I. No. The replacement of the temperature transmitter 1G33-N003 vith four single units is classified as a nonsafety-related change. .All power for this design change is provided by a non-lE supply. These devices are for temperature monitoring only and do not isolate or trip any components in case of equipment failure. Therefore the l probability of occurernce or the consequences of an accident or . malfunction of equipment important to safety previously evaluated in the USAR is not increased. II. No. These instruments are passive and the possibility for an accident or malfunction of a different type other than evaluated previously in the USAR is not created. III. No.. The installation of the new transmitters for the RVCU system vill not impact nor reduce any safety margin as defined in the Technical Specifications. l
'AttIchm:nt 3 PY-CEI/NRR-1141 L Page 280 of 285 '
;3E No.: 89-205 dource Document: SSCR CR 89-119 Lescription of Change This change incorporates changes associated with modification to the Control Room HVAC (H25) System into the Fire Protection Safe Shutdown Capabilities Report (SSCR). These changes consist of listing the new circuits in the SSCR.vhich originated when the M25 modification was installed. The modification requires that two chlorine detectors must -
identify the presence of chlorine before a HVAC isolation takes place. Summary I. No. . This change just incorporates documentation of the M25 system modification into the SSCR. The change does not affect the 10 CFR'50 Appendix R Basis. Hence, tne probability of occurrence or the consequence of a fire is not significantly increased. II. No. This change does not create a new type of-accident or malfunctlon of a different type than previously evaluated in tie USAR. i III. No' . There is no relationship betvcen the change and the basis for any < Technical Specification. Thereford, the proposed change does not reduce the margin of safety as described in Technical Specification.
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Attachm:nt 3 PY-CEI/NRR-1141 L ! Page 281 of 285-SE No.: 89-206 , Source Document: SSCR 89-096 Desc'ription of Change-This evaluation analyzes changes to the Fire Protection Safe Shutdown Capabilities Report (SSCR). The changes are editorial in nature. Summary
~
I. No, The changes are editorial in nature and improve the readability, and b quality of the.SSCR. No systems or components are altered. , lc .Therefore, the probability of occurrence or consequences of an
;> accident or malfunction of equipment previously evaluated is not increased.
II. No. Changes are editorial in nature.. Therefore, the' chance of an
, accident oc malfunction of a different type than any evaluated' ,
t previously in the USAR does not exist. < III.'No. Changes are editorial in nature and vill not affect the operability. ; or availability of any system described in the Technical' Specifications. This change vill not reduce the margin of safety as defined in the basis for any Technical Specification. L f t i , .? r )
, _-m Attachm:nt 3 p
PY-CEI/NRR-ll41 L < d Page 282 of 285 L SE Nor 89'-207 ! source Document: DCP 87-390A, Rev. O ! Description of Change The design change modifies the Intermediate Building Ventilation (M33) System duct run to accommodate the addition of a bulk oil / grease storage - area. This modification extends two exhaust registers to within 12 inches of the floor and adds fire dampers in the duct for fire protection requirements for the addition of the Bulk 011 Storage Room. Summary
- r F
I. No. USAR Section 9.4.7.2 states that the Intermediate Building
, Ventilation System is nonsafety-related, non-seismic and is not required to safely shutdown the plant. The H33 oystem function vill not be adversely affected by this design change. Therefore, the probability of occurrence or the consequences of an accident or .
malfunction of equipment important to safety previously_ evaluated in l the USAR vill not be increased. II. No. As described in Item I above, H33 system operation has not been , altered. Hence, the possibility of an accident or malfunction'of a different type than previously evaluated is not created. III.~No. This change involves only changes to the Intermediate. Building Ventilation System which is not covered within Technical r-Specifications.. Therefore, the margin of safety within any Technical Specifications is not reduced. t r3
n -_ 4 Attachm:nt 3 PY-CEI/NRR-1141 L Page 283 of 285 SE No.: 89-208
% Source Document: DCP 88-363, Rev. O Description of Change This design change modifies the Circulating Vater Pumphouse Ventilation -
(H45) System. An Allen Bradley Programmable Logic Controller is used to [ provide a four stage temperature control by manipulating the exhaust fans, the fcn inlet dampers and the vall louvers. Summary I. No. This change modifies the M45 operating controls.as described in USAR Section 9.4.12.2.2. The installation of the controller vill reduce the mechanical forced air operation and cycling. The system vill
-rely more upon natural circulation to maintain building temperature.
However, the overall H45 system design function has not been , altered. Therefore, the probability of the occurrence or , consequences of an accident or malfunction of equipment previotely evaluated is not increased. II. No. This change vill not increase the possibility for an accident or malfunction of a different type than any evaluation previously in the USAR. .The M45 system is not required for safe shutdown of the plant. III. No. The M45 system controls are not described in the Technical Specifications. Therefore, the margin of safety as defined in the
~ bases-of the-Technical Specification is not reduced.
AttcchmInt 3 P7-CEI/NRR-1141 L-Page 284 of 285
-SE No.: 89-209 I
= Source Document SCR 89-0005-T
, Description of Change This temporary setpoint change request eliminat*s a continuous alarm in the Control Room due to Safety Relief Valve (SRV) veepage. This temporary change request increases the SRV leakage high temperature alarm setpoint on 1B21-R0614 recorder. Summary
.I. No. Incorporation of this temporary change raises the nonsafety high-I' temperature alarm setpoint on the 1B21-R0614 recorder. The recorder vill continue to track SRV discharge pipe temperature for all nineteen channels. . Increasing the leakage high temperature alarm setpoint is considered an' improvement to plant operation,' as it eliminates the distraction of a continuous alarm due to SRV veepage.
- Safety-related SRV tailpipe pressure indication is unaffected by this temporary: change. Therefore, the possibility of an accident or malfunction of a different type than any. evaluated previously in the USAR is not created.
II. No. See Item I. III. No. This temporary change vill not affect the operability or -f availability of any system described in the Technical Specifications.- Therefore, margin of safety is not reduced by this temporary change. m-
Attachment'3 PY-CEf/NRR-1141 L Page 285 of 285
.i SE No. :: 89-210 SourJi Document:- DCN 2870, Rev. O Description of Change This drawing change is' editorial in nature. It corrects panel locations
- ," of electronic equipment on P&ID D-302-621, Emergency Closed Cooling (P42) System.
Summary L
-. I. No. This change is editorial in nature. It corrects the location of electronic equipment only. It-does not alter the design or
' operation of_the P42l system. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated is'not increasad.
II. No. Changes are editorial only. The chance of an accident or malfunction of a different type than any evaluated previously in the JUSAR does not exist. III'. No. This change vill not affect-the operability or availability of any system- described in the. Technical Specifications. Therefore, the margin of safety as defineu in the basis for any Technical Specification is not reduced. I
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