ML20244C395
| ML20244C395 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 04/26/1989 |
| From: | Karner D ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 161-01868-DBK, 161-1868-DBK, TAC-72713, NUDOCS 8906140263 | |
| Download: ML20244C395 (176) | |
Text
-
tac 72713 b
Arizona Nuclear Power Project P.O. box $2034 e PHOENIX. ARIZONA 85072-2034 161-01868-DBK/GEC 1
April 26, 1989 l
I Docket Nos. STN 50-528/529/530 l
U. S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station F1-137 Washington, D. C.
20555 P
Reference:
(A) Letter from T. L. Chan, NRC, to D. B. Karner, APS, dated March 29, 1989.
Subject:
)
Request for Additional Information -
10CFR 50.59 Annual Report for 1987
Dear Sirs:
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 1, 2, and 3 Response to Request for Additional Information -
10CFR 50.59 Annual Report for 1987-File:
89-056-026 In response to your request, please find attached the entire 10CFR 50.59 evaluation packages identified in Reference (A).
The attachment includes a j
cross-reference between the change numbers identified in Reference (A) and the identifying numbers found on each 10CFR 50.59 evaluation package.
If you have any questions, please contact A.
C.
Rogers of my staff at (602) 371-4041.
Very truly yours, D. B. Karner l
Executive Vice President RAB/GEC/jle l
Attachment ec:
T. L. Chan J. B. Martin T. J. l'olich
(
8906140263 890426 g\\
R ADOCK 05000528 PDC
l.
l ATTACHMENT 10CFR 50.59
~ Change Number Identifying Number / Description 17 SARCN 2183-27 SARCN 2242 33 SARCN for EER 85-SQ-109 (SARCN 2258) 36 SARCN 2262 1
44 FSAR Change to Section 1.8 (SARCN 2273) 58 Deviation to Fire Protection Criteria in FSAR Section 9B.2.12 and 9B.2.15 (SARCN 2289) i 73 SARCN 2311 81 SARCN 2322 87 Add Deviation to 10CFR 50, Appendix R, Section III.G.2 to FSAR Section 9B.2 (SARCN 2352) 91 Change to FSAR 6.2.4 and 6.2.6 (SARCN 2359) 146 DCP 10M-RC-147 l
150 DCP 10J, 20J. 30J-AF-086 l
1 m_____._________
Summary of Safety Evaluation 1'
l~
The. use of this code case by ANPP does not introduce an unreviewed-safety question since the code case has been approved for use by the NRC and has been accepted in Reg. Guide 1.84'.
(14).
Description i
The purpose of this change was to modify the QA program described in FSAR Sections 17.lA and 17.2 to incorporate administrative control and l
organization changes.
These changes did not reduce licensing commitments and were transmitted to the NRC in Amendment 17 of the FSAR.
Summary of Safety Evaluation This change did not introduce an unreviewed safety question.
The change involves only the description of the QA program, and does not impact important to safety equipment.
(15)
Description FSAR Tables 8.3-3 and 9.1-1 were revised to reflect the correct data for the fuel pool cooling pumps. The changes reflect actual pump data
.from vendor information and nameplate inspections.
This change was transmitted to the NRC in the USAR, Rev. O.
Summarv of Seferv Evaluation This change did not introduce an unrevDPwed safety question. The fuel pool cooling pumps still deliver their rarr d flow and the increased load of the pumps does not significantly reduce the diesel generator margin. The operation of the pumps is not affected.
(16)
Description o
FSAR Section 11.5.2.1.1.7.2 and Table 11.5-1 were revised to reflect the replacement of the moving filter mechanisms of radiation monitors J-SQN-RE-8, 14, 13A and 133 with fixed filter mechanisms.
This chan58 was transmitted to the NRC in the USAR, Rev. O.
Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. The monitors were modified due to deficiencies in the original design, and the modification sicplified operation and maintenance of the monitors.
In addition, the monitors are non-safety related equipment.
)((17)
Descriorien This change revised FSAR Section 2.2. 18.III..D.3.4 and Table 6.4-3 to note that the PVNCS control room does not have chlorine detectors and that CRVIAS can only be initiated manually.
In addition. the reference that PVNGS has a Type B control room per Reg. Guide 1.78 was deleted.
This change was transmitted to the NRC in the USAR. Rev. O. -
[ '..
Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. The change reflects current design and has no impact on any TSAR accident analyses.
1 (18)
Description FSAR Sections 3.2 and 12.5.2.2.7 were revised to reflect the replacement of the Radiation Exposure and Management (REM) system by the Radiological Records and Access Control (RRAC) System. This change was transmitted to the NRC in the USAR, Rev. O.
Summarv of Saferv Evaluation This change did not introduce an unreviewed safety question.
The REM /RRAC system is not important to safety and operation of the system will not impact any safety related system. The REM /RRAC system is also not taken credit for in any safety analyses.
l l4 (19)
Description FSAR Section 8.3.1.4.1.1 is revised to take exception to the separation 1
requirements of Reg. Cuide 1.75 for the new fuel handling crane and containment refueling machine.
Similar exceptions have been accepted I
by the NRC as documented in SER Supplement 7.
This chanSe was transmitted to the NRC in the USAR, Rev. O.
Sumenry of Saferv Evaluation This change did not introduce an unreviewed safety question. The l
change documents the as-built condition of the plant.
The cranes are e
used infrequently and operate only during short periods. When not in use, they are deenergized. Since the cranes are locally controlled, I
any malfunction would be detected by the operator and appropriate action initiated.
(20)
Description FSAR Section 9.5.4.2.1 was revised to provide additional clarification regarding the diesel fuel oil storage tank capacity, to eliminate continued confusion relating to the volume provided for periodic testing. This change was transmitted to the NRC in the USAR, Rev. O.
j Summarv of Safetv Evaluation This change did not introduce an unreviewed safety question.
Clarification was provided regarding the diesel fuel oil storage tank capacity, but the system or its mode of operation was not altered in I
any way.
1 6-
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yg FILE: 97-0 39 'lCY PAGE / OF L_
10CFR50.59 REVIEW AND EVALUATION t
ACTION UNDER REVIEW: gx7#dA/ o/M 3 REVISION://AP DESCRIPTION OF PROPOSED CHANGE:g4# /o de elo ao e e-/ /o suo /r,/ A a /,
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DOES THE PROPOSED CHANGE:
1.
MAKE CHANGES IN THE FACILITY? ~
YES )( NO 2.
MAKE CHANGES IN PROCEDURES AS THEY ARE DESCRIBED IN THE FSAR?
YES NO '>C 3.
INVOLVE TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR?
YES NO X 4
INYOLVE ANY OTHER CHALLENGES TO NUCLEAR SAFETY FOR PVNGS?
YES NO X 5.
REQUIRE A' CHANGE TO THE TECHNICAL SPECIFICATIONS?
-~~
YES N01 10CFR50.59 EVALUATION (Provide Response Justification with References) 6.
WILL THE PROBABILITY OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES N01 THE FSAR BE INCREASED?
7.
WILL THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES N0 1 THE FSAR BE INCREASED 7 8.
WILL THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES Nel TO SAFETY BE INCREASED?
9.
WILL THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES N0 1 TO SAFETY BE INCREASED?
i
- 10. WILL THE POSSIBILITY OF AN ACCIDENT OF A DIFFERENT TYPE THAN YES N0_ X ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
II. WILL THE POSSIBILITY OF A MALFUNCTIONING OF A DIFFERENT TYPE YES N0 1 THAN ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 12. WILL THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY YES NO..X TECHNICAL SPECIFICATION BE REDUCED?
X ANY ANSWER TO QUESTIONS I THROUGH 4 "YES", THEN A 10CFR50.59 EVALUATION IS REQUIRED. FSAR CHANGE REQUEST PER PROCEDURE 5N404.01.00 MA REQUIRED.
ANSWER 5 IS "YES" THEN TECHNICAL SPECIFICATION CHANGE REQUEST PE PROCEDURE 5N404.01.00 AND NRC APPROYAL REQUIRED PRIOR TO I ANY ANSWER TO QUESTIONS 6 THROUGH 12 "YES" THEN AN UNREVIEWED QUESTION IS IDENTIFIED.
PROCEED 10 PROCEDURE 7N407.03.00 PRIOR TO IMPLEMENTATION.
ALL ANSWERS 6 THRU 12 ARE "NO" RECO:BIEND ACTION APPROVAL.
ALL ANSWERS 1 THROUGH 5 ARE "N0", NO 10CFR50.59 EVALUATION REQUIRED, RECO:f!END ACTION APPROVAL.
I verify that the above revicideraluation is adequate and accurate and that at least one of the undersy;ned has received required training.
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~ II.III.' D. 3. 4 CONTROL ROOM HABITABILITY REQUIREMENTS Position In accordance with Task Action Plan item III.D.3.4 and control room habitability, licensees shall assure that control room.
operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19, " Control Room,"
of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50).
PVNGS Evaluation
.\\
Potential hazards in the vicinity of the site are discussed in FSAR Section 2.2.
The operators in the control room are ade-quately protected ffom these hazards and the release of.tadio-
/
active gases as discussed in FSAR Section 6.4.
The required information provided below is in the format suggested by to NUREG 0737 Section III.D.3.4.
INFORMATION REQUIRED FOR CONTROL-ROOM HABITABILITY EVALU (1)
Control-room mode of operation:
automatic filtered recir-culation with filtered makeup for pressurization for radio-ha s w e.1 logical accident isolation.
':t=tig filtered recircula-tion without makeup for chemical acci' dent isolation.
Manual smoke removal mode (operators alerted by smoke detector).
..s 13.III.D.3.4-1
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NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES mostly liquified hydrocarbons such as LPG, propane, butane and a few other materials such as methyl chloride and vinyl j
The other hazardous material category could include non-flammable compessed gases other than ammonia and chlorine l
(such as hydrogen chloride or inert gases; nitrogen, argon, L
etc.) flanmable or combustible liquids, flammable solids, and corrosive materials.
The latter three categories include a large number of chemicals and consumer goods which, because of small package-size or low volatility, are no hazard to PVNGS 2
at the 4.5 mile distance.
As indicated in section 2.2.2.2.2, no poisonous gases, Class A explosives, radioactive materials, acrylonitrile or hydrofluoric acid was carried past the plant.
From the above discussion, studies of hazardous material storage and transport (22, 23, 24, 25) and from a review of i
1973 to 1977 records of spills reported to the Office of
, Hazardous Materials, Department of Transportation, the chemi-
'--cals listed in table 2.2-3 were identified as potentially shipped past PVNGS and a potential toxic chemical hazard for further investigation'.
This list includes materials shipped as compressed liquified gases and involved in two or more rail accidents that lead to a spill during the 5-year period 1973 to 1977.
It should be noted that of these materials, only LPG (propane or butane) and anhydrous ammonia were involved in spills in the five-state region; Arizona, New Mexico, Utah, Nevada, and California.
It is expected that most of the 185 cars of flammable compressed gases shipped past the plant in 1978 were LPG since this is by far the largest commodity in this category shipped.(25)
The Regulatory Guide 1.78 allowable weight of the commodities in table 2.2-3 was determined conservatively for a Type C con-trol rooN""C ^ '.- - M: : n.. - ; ___. M th an air exchange rate of 0.45 h~l, type G dispersion and a 4.2 mile l
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Summary of Safetv Evaluation i
This change did not introduce an unreviewed, safety question. The design basis accident applicable for RU-1 is a
containment overpressurization.
RU-1 is not taken credit for in the accident analysis since.the CIAS valves perform the safety related function in this situation.
(25) peserietion FSAR Section 6.4.5, 6.4.7, 6.5.1.4, and Table 9A-1 were revised to reflect the approval by the NRC of the 1980 revision to ANSI N509, concerning charcoal filters. Reg. Cuide 1.52 has not yet been revised and still references ANSI N509-1976.
This change was transmitted to the NRC in the USAR, Rev. O.
Summary of Safety Evaluation This change did not introduce an unreviewed safety question'. The inservice test requirements for the charcoal are not changed and the decontamination efficiencies required by the FSAR will still be met.
(26) peserietion FSAR Table 8.3-6 describes Class 1E DC System Loads, however, the format,of the FSAR table is not consistent with the format of the Bechtel Calculation that identifies the data used in the table.
Therefore, FSAR Table 8.3-6 was revised to make the format consistent with the calculation, and to incorporate results that were originally omitted. Th41 change was transmitted to the NRC in the USAR, Rev. O.
Summarv of Safetv Evaluation 1his chance did not introduce an unreviewed safety question.
The changes were editorial in nature and did not impact the ojeration or functional requirements of the Class lE DC power system.
}
(27)
Description FSAR Sections 1.8, 5.2 and 11.5 were revised to document the correct location of the RU 1 sample point, ANPP's position on Reg. Cuide 1.45 (taking exception to the capability of RU-l to quantify leakage) and ANPP's methods for leakage detection.
This change was transmitted to the NRC in the USAR, Rev. O.
Summary of Safetv Evaluation This change did not introduce an unreviewed safety question. RU-l will alarm on abnormally high radiation levels and the control room operators can respond appropriately to increasing trends. The function of the detection system remains the same, and other cethods exist to quantify the leakage.
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PAGE OFD l
10CFR50.59 REVIEW AND EVALUATION ACTION UNDER REVIEW:
6 Mkh UD REVISION:
f DESCRIPTION OF PROPOSED CHANGEb8 #YWL. M PO fmA br 0
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IK1A 11
- 3
. C RbM O 10CFR50.59 REVIEW DOES THE PROPOSED CHANGE:
1.
MAKE CHANGES IN THE FACILITY?
YES / NO 2.
MAKE CHANGES IN PROCEDURES AS THEY ARE DESCRIBED IN THE FSAR? YES_ / NO 3.
INVOLVE TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR?
YES NO t/
4 INVOLVE ANY OTER CHALLENGES TO NUCLEAR SAFETY FOR PVNGS?
YES NO 1/
5.
REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS?
YES N0 i/
10CFR50.59 EVALUATION'{'hrovideResponseJustificationwithReferences) 6.
WILL THE PROBABILITY OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES N0 t/
THE FSAR BE INCREASED?
7.
WILL THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED INYES N0 t/
THE FSAR BE INCREASED?
8.
WILL THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES NO 1/
TO SAFETY BE INCREASED?
9.
WILL THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES NO /
TO SAFETY BE INCREASED?
10.
WILL THE POSSIBILITY OF AN ACCIDENT OF A DIFFERENT TYPE THAN YES NO /
ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 11. WILL THE POSSIBILITY OF A MALFUNCTIONING OF A DIFFERENT TYPE YES NO /
THAN ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
f
- 12. WILL THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY YES NO V TECHNICAL SPECIFICATION BE REDUCED?
/ ANY ANSWER TC QUESTIONS 1 THROUGH 4 "YES", THEN A 10CFR50.59 EVALUATIO IS REQUIRED. FSAR CHANGE REQUEST PER PROCEDURE 5N404.01.00 MAY ALSO BE REQUIRED.
ANSEER 5 IS "YES", THEM TECHNICAL SPECIFICATION CHANGE REQUEST PER PROCEDURE 5N404.01.00 AND NRC APPROVAL REQUIRED PRIOR TO IMPLEMENTA ANY ANSWER TO QUESTIONS 6 THROUGH 12 "YES" THEN AN UNREVIEWED SAFETY QUESTION IS IDENTIFIED. PROCEED TO~ PROCEDURE 7N407.03.00 PRIOR TO IMPLEMENTATION.
ALL ANSWERS 6 THRU 12 ARE "NO" RECOSDIEND ACTION APPROVAL.
ALL ANSWERS 1 THROUGH 5 ARE "N0", NO 10CFR50.59 EVALUATION REQUIRED, RECOSDIEND ACTION APPROVAL.
I verify that the above review / evaluation is adequate and accurate and that at 1 cast one of the undersigned has received y
required training.
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PAGE OFD 10CFR50.59 l
REVIEW AND EVALUATION i.
1 ACTION UNDER REVIEW:
b@D TlM1 REVISION:
I
- DESCRIPTION OF PROPOSED CHANGEk()hVY\\EL NPO PnA -
0 lh,C D-b bl RbH D
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10CFR50.59 REVIEW DOES THE PROPOSED CHANGE:
1.
MAKE CHANGES IN THE FACILITY?
YES / N0
?
'2.
MAKE CHANGES IN PROCEDURES AS THEY ARE DESCRIBED IN THE FSAR?YES / NO
{
3.
INVOLVE TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR?
YES NO v 4
INVOLVE ANY OTHER CHALLENGES TO NUCLEAR SAFETY FOR PVNGS?
YES N0 t/
5.
REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS?
YES NO g/
10CfR50.59 EVALUATION { Provide Response Justification with References) 6.
WILL THE PROBABILITY OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES NO /
THE FSAR BE INCREASED?
7.
WILL THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED INYES NOt/
THE FSAR BE INCREASED?
8.
WILL THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANTYES Nog TO SAFETY BE' INCREASED?
9.
WILL THE CONSEQUENCES OF A MALFUNCTION.0F EQUIPMENT IMPORTANT YES NO /
TO SAFETY BE INCREASED?
- 10. WILL THE POSSIBILITY OF AN ACCIDENT OF A DIFFERENT TYPE THAN YES NO /
ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 11. WILL THE POSSIBILITY OF A MALFUNCTIONING OF A DIFFERENT TYPE YES NO /
THAN ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
_j
- 12. WILL THE MARGIN OF SAFETY AS DEFINED IN.THE BASIS FOR ANY YES NO V TECHNICAL SPECIFICATION BE REDUCED?
/ ANY ANSWER TO QUESTIONS 1 THROUGH 4 "YES", THEN A 10CFR50.59 EVALUATIO IS REQUIRED. FSAR CHANGE REQUEST PER PROCEDURE 5N404.01.00 MAY ALSO BE REQUIRED.
ANSWER 5 IS "YES", THEN TECHNICAL SPECIFICATION CHANGE REQUEST PER PROCEDURE 5N404.01.00 AND NRC APPROVAL REQUIRF.D PRIOR TO IMPLEMENTAT ANY ANSWER TO QUESTIONS 6 THROUGH 12 "YES" THEN AN UNREVIEWED SAFETY QUESTION IS IDENTIFIED. PROCEED TO' PROCEDURE 7N407.03.00 PRIOR TO IMPLEMENTATION.
[ ALL ANSWERS 6 THRU 12 ARE "NO" RECO.'D!END ACTION APPR ALL ANSWERS 1 THROUGH 5 ARE "N0", NO 10CFR50.59 EVALUATION REQUIRED, RECO.'DIEND ACTION APPROVAL.
I verify that the above review / evaluation is adequate and accurate and that at least one of the undersigned has received E.
required training.
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CONFORMANCE TO NRC REGULATORY GUIDES Since severe sensitization is avoided, testing to determine susceptibility to intergranular attack is not performed.
6 REGULATORY GUIDE 1.45:
Reactor Coolant Pressure Boundary Leakage Detection Systems (Revision 0, May 1973)
RESPONSE
e N N orYDe-The position o Regula ide 1.45 is accepted,g(-refer too hfk8d N.b 5-)v Ylso see CESSAR Section 1.8.
y REGULATORY GUIDE 1.46:
Protection Against Pipe Whip Inside Containment (Revision 0, May, 1973)
RESPONSE
Except as discussed below, protection against pipe whip inside the containment complies with Regulatory Guide 1.46.
A.
Position C.1.b of the Regulatory Guide:
Intermediate break locations between terminal ends are postulated to occur at weld joints where the piping incorporates a fitting, valve, or welded attachment where the following are met (as specified in NRC Branch 5
Technical Position MEB 3-1):
5l (1) the stress range S exceeds 2.4 S,, where S n
m is the design stress intensity as defined in Section III of the ASME Code, or (2) the stress range S as calculated by equation (10) n
[
of Paragraph NB-3653 exceeds 2.4 S and the E
stresses computed by equations (12) and (13) of 5
Paragraph NB-3653 are greater than 2.4 S,,
or (3)
If fatigue analysis is performed, any intermediate locction between terminal ends where the cumulative Amendment 5 1.8-30 August 1981
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PUNGS7FSAR l(
INTEGRITY OF REACTOR-COOLANT PRESSURE BOUNDARY 5.2.3.4.2 Control of Welding 1
5.2.3.4.2.1' Avoidance of Hot Crackinn.
A.
Components in C-E Scope of Supply Refer to CESSAR Section 5.2.3.4.2.1-A.
B.
Components Not in C-E Scope of Supply In order to preclude microfissuring'in austenitic stainless steel, PVNGS design is consistent with the recommendations of Regulatory Guide 1.31 except as noted in section 1.8.
5.2.4 INSERVICE INSPECTION AND TESTING OF REACTOR COOLANT PRESSURE BOUNDARY 2
Details of.the inservice inspection program are included in 15 section 6.6, and Section 3/4.0 of the Technical Specifications.
Accessibility of inspection areas is discussed in CESSAR i
Section 5.2.4.1.
5.2.4.1.
DELETED 5.2.5 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS Means for the detection of leakage from the reactor coolant pressure boundary.are provided to alert operators to the exis-tence of leakage above acceptable limits, which may indicate an unsafe condition for the facility.
The leakage detection systems are sufficiently diverse and sensitive to meet the Y
criteria of Regulatory Guide 1.45 for le ks from identified and unidentified sources.
M h*
N !'
5.2.5.1 Leakace Detection Methods 5.2.5.1.1 Unidentified Leakage The four methods employed to detect unidentified leakage are presented in the following sections.
12 Amendment 15 5.2-136 April 1986
'*l'
'Ac'LYG.R.I 5 g/t/
[
7 INTEGRITY OF REACTOR COOLANT PRESSURE BOUNDARY
'e Safety injection line pressure High-pressure safety injection header pressure e
[
Low-pressure safety injection header pressure e
5.2.5'.2.3 Leakage Detection Conversion to Leakage Equivalent The procedures necessary for 'each leak. detection method to be converted to a common leakage equivalent are indicated below.
5.2.5.2.3.1 Containment-Radioactive Gas and Air Particulate fahighactivityalarmkh Monitoring., dU O..a @ Nion* W ***e '
- reasM!I w
t%
W'(AAlfaSitutA Wht:r bumsvto C,oe@ct,d h /Ldi9.
[-
A.
Reiords particulate and gaseotrs-activity measurements at equal time intervals.
B.
Determines the activity increases during these intervals.
C.
Estimates RCS activity level based on previous operat-ing history.
D.
Estimates the coolant leakage rate from the radioac-tivity increases.
This is accomplished by comparing the measured increases to a curve showing abnormal primary coolant leakage as a function of particulate and gaseous activity variation.
Estimated leakage and measured particulate and gaseous activi-ties then are plotted to test the consistency of the data and to estimate the time elapsed since the beginning of the leak.
~
Nonconstant activity increases mean either:
Containment atmosphere activity is approaching steady o
state
\\
The abnormal leakage flow is not constant.
e
\\.
12 Amendment 12 5.2-142 February 1984
_ _ _ _ ______ _ _______________ ___ _____._________________ _____ _______ _______ __________ _ __ _____j
g-l *..
A0 YstI kW PUNGS FSAR PROCESS AND EFFLUENT RADIOLOGICAL 4
MONITORING AND SAMPLING SYSTEMS.
y B.
Provide for early detection of radioactive leakage-into normally. nonradioactive systems, including b
' primary-to-secondary leakage, primary-to-atmosphere leakage, and process system' leakage into normally non-radioactive systems.
-Specifically}ncluded is the I
capability of both 'he containment building gaseous t
i4 channel and the containment. building particulate channel each to detect independently an increase in the reactor coolant system-to-containment atmosphere leak rate ef 1 galf=in "ithin 1 heur?'as two of the methods of leak detection required to follow the g >X.c.yd u d i 0
-recommendation of NRC Regulatory Guide 1.45 e
.EJ4t4 crn I.5 C.
Provide continuous. remote indication and recording of airborne radioactive contamination in the form of particulate.and iodines in areas where-personnel-normally have access, except in areas where the poten-tial for airborne activity releases is negligible, in order to follow the recommendations of NRC Regulatory.
Guide 8.8 for control of occupational exposure to radiation.
l' 11.5.1.1.2 Effluent Monitoring System The effluent monitoring system is designed to perform the fol-lowing functions in order to meet the requirements of 10CFR20, 10CFR50, and follow the recommendations of Regulatory Guide 1.21 during normal operations, including anticipated operational occurrences:
A.
Provide continuous representative sampling, monitoring, recording, and indication of gaseous radioactivity levels, and, as a minimum, continuous representative sampling of particulate and iodine radioactivity levels along principal effluent discharge paths.
,.1 11 C.?
M#2 l 7
~'~
PROCESS AND EFFLUENT RADIOLOGICAL MONITORING AND SAMPLING SYSTEMS s
11.5.2.1.3.13 Containment Building Atmosphere Mcnitor, (CBB) 12 Channel "B" (XJ-SQB-RU-01).
The containment building atmo-
~
sphere is continuously monitored for particulate, iodine, and gaseous activity.
The sample is drawn from the containment building in a closed system, is monitored outside the contain-ment, and then is returned to the containment building atmosphere after it passes through the samplers.
The particulate and gaseous channels serve as two methods of RCPB leak detection in accordance with Regulatory Guide 1.45p.LLK0Ap CW This monitor is designated seismic Category I.
UN M II In addition to the three radiation channels a hygrometer which measures dewpoint temperature is provided for the CB-B monitor.
The hygrometer is microcomputer-controlled and is configured to be an extra channel of the monitor.
The hygrometer channel 12 measures dewpoint temperature over the range of 0-200F.
Indi-cated dewpoint is accurate to within +.4F with respect to actual dewpoint of the sampled air at a rate of change of dew-point up to +3F/sec.
This channel can be used as a method of RCPB leak detection.
In order to obtain a representative sample of containment air, the sample line inlet is located on the ope,ra g egel between two of the normal cooling units intakes.
This location also facilitates RCPB leak detection by these monitors.
Available monitor sensitivities allow the particulate and I-131 channels of the monitor to detect maximum permissible concen-12 trations allowed by 10CFR20 in the containment building within one hour for Cs-137 and within eight hours for I-131.
l The CB-B monitor is located just outside the containment build-ing.
It samples the containment atmosphere through piping penetrations.
Isolation valves at the monitor automatically l2
.Lsed b g A,n cm +>v loc /-o" uulvedum m M e u//5~3, g
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u 1984 11.5-39 Amendment 12 L -__ ___- _-__bruarv
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l4, Ol l ~ (32)
== Description:== FSAR Section 1.8 'was. revised with respect to Reg. Cuide 1.52 to allow .the use of silicone type sealants for sealing electrical and piping penetrations in-the_ fuel building HVAC (HF) and control building HVAC (HJ)' essential air filtration units. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The use of the, silicone sealant will not affect the ability of the HF and HJ ~ systems to perform their safety related function. Any degradation of the sealant would not result in a leakage. path for contamination into the control room enevelope. I. (33) Description The range of the main steam support structure effluent monitors described in FSAR Table 11.5-1 is changed from 10-3 to 10' R/hr to 10' to. 105 mR/h. This is consistent with Reg.~ Cuide 1.97 requirements. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question..The change documents.the.as-built' condition of the monitors, which still meet the Reg. Guide 1.97 requirements. (34) Description FSAR Section 1.1 is revised to reflect the transfer of 5.7% of undivided ownership from Salt River Project (SRP) to the Los Angeles Department of Water and Power. Pursuant to an agreement dated 8/18/77, this ownership was transferred when PVNGS Unit I was placed in commercial service. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. The ownership does not impact the operation of the plant or the function of safety related equipment. (35) Description FSAR Table 11.5-1 was revised as follows: The listing of the gas charcoal alarm setpoint values for monitors RU-12, RU 141, RU-142, l RU-143, RU 144. RU-145, and RU 146 was, deleted and replaced by a footnote referencing the tech spec. The tech spec requires that the setpoints be determined in accordance with the Offsi:e Dose Calculation Manual (ODCM) and could change periodically as the ODCM is revised. In 995$ r V (o FnE # 67-O W lo/ 10CFF.50. 59 PAGE / OF. PlVIEW AND EVALUATION ACTION UNDER FIVIEW: Q.- O-b FC 95-5 O - to 9 c-FRISION: DESCRI?!!ON OF PROPOSID CHANGE: A> Ar <w' ^ b ~a'<=' e c M x A La L a m n n L~ u-,~s n'Lf-G 9 E' / d O 10CFR50.59 F2 VIEW DOES TE PROPOSED CEANGE: 1. }t03 CHANGES IN TEE FACILIIT AS IT'S DISCFl3ED IN TEE FSAR? 2. 1102 CHANGES IN PROCDUFIS AS TEIT ARI DESCRIED IN THE FSAR? ES X NO 3. DiVOLVE FISTS OR II?IFl.".:.NTS NOT DESCRI3ED IN THE FSAR7 TIS NO2 4 DiVOLVE A:C On2R CEALLINGES TO NUCIIAR S/JE~T FOR PVNOS? TIS NO g 5. RIQUI?I A CHANGE TO TE TECDilCAL SPECIFICAHONS? TIS NO x TIS S NO[ 10CTR50.59 IVALUATION (? ovide Response Justification with References) 6. VII.L THE FRO 3AIII.ITY OF AN ACCIDDIT I?2VIOUSLT IVALUATED IN FSAR E INC?l.ASD? TES NO V 7. VH1 THE CONSEQUDiCES OF AN ACCIDDC PREVIOUSLT EVEUAT w FSAR E INC?mED? TIS NO E 8. VIII TE PROSA31LITT OF A MAIJUNCn0N OF IQUI?}21C IMPORTEC T S;5ITY E.DiCFl.ASD? TES NO 1 9. VIII TE CONSEQUD'CF.S OF A MALFUNCTION OF EQUIPMDC IMPORT;2C TO SIJE!! 3I DiCFJ.iSD? TIS NO 1
- 10. WILL THE ?CSSILILI"T OF AN ACCIDDIT OF A DIFFFJZ2C TYPE T
??.IVIOUSLT IVAI.UATID IN THE FSAR BE CPIATD? TES NO L
- 11. WII.L TE POSSIBILITY OF A MAI. FUNCTIONING OF A DIFFIFINT T AST ?REVIOUSLT EVALUATED IN TE FSAR 3E CFJJJED?
TES NO j' 12. WILL n2 M;Jari 0F S;JI T AS DEFINED ;N D2 3 ASIS FOR ANY TECENICAL S?ICIFICATION EI FIDUCD? YES NO X M RW ANSWFJ. TO QUESTIONS 1 THROUGE 4 "TES", TEN A 10CFF.50.59 IVALU E QUIRID. ANSWFJ. 5 IS "ES", THEN TECENICAL SPECIFICA20S C M 5N404.01.00 AND NRC APPROVAL FIQUIED ??.IOR TO IMPLEM72 CATION. IliY ANSWF.R TO QUESTIONS 6 TERCUOE 12 "TES", TEDi AN UNF.IVIEWE IS IDECIFIID. ALL ANS'aTJ.S 6 TEROUGH 12 AFI "NO", PIC010END ACHON IJFROV/2. PROCEED TO K ALL ANSWERS 1 THROUGH 5 AF2 "NO", No 100FF.50.39 EVAI.UAn08 EQULy2D ACTION IJ?ROVAL. , EC0ygegg I verify tha: the above review / evaluation is adecus:e cud accurate and tha 1 at leas: c:e I cf the undersigned has received required
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addition, the particulate.and iodine channel monitors listing for RU-142. RU-143, RU-144 and RU-146 alarm setpoint values was deleted.and replaced by a footnote referencing ALARA limits._ The values for these monitors are defined by ALARA limits, which change due to varying radiation levels in the monitor areas. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The changes makes the FSAR consistent with tech spec requirements and ALARA considerations. (36) Description The Onsite Power System descriptions of FSAR Section B were revised to reflect-as built conditions. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safety Evaluation This change did not introduce un unreviewed safe ty question. The changes do not alter any important to safety equipment or assumptions considered in the FSAR accident analyses. F. (37) Description FSAR Section 9A was revised to document ANPP's exception to Reg. Cuide 1.140, allowing the use of silicone sealant on containment preaccess filter units. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safetv Evaluation This change did not* introduce an unreviewed safety question. The pre-access filter units provide no safety related function and their failure or decreased performance cannet initiate an accident or result in the release of an offsite dose. (38) Descriorion FSAR Section 6.4.4.2 was revised to accurately reflect sources of chlorine gas onsite. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Seferv Evaluation This change did not introduce an unreviewed safety question. For the most conse rvative chlorine gas release, the maximum concentration of chlorine gas at the control room air intake structure in below the toxicity limits specified in Reg. Cuide 1.78. The source of the chlorine is the electrolytic cells in the water reclamation facility and as such no safety related systems are impacted. - _ _ _ _ _ _ - _ -
c. U.GC 22W/3 10CTR50.59 Review and Evaluation ACilDe t*.tER REVIDJs s A n u s n.i 2.o 6 L IOrISIGN: n t DECCTIIPTIO4 0F PRTO5ED DIANCE: - U %lOs D5 AA1 ~ lAA. f)3 a w m
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o r. sa 10CTR50.'.,9 REVICJ (Previce Peferences en Resocnse Justification Page) to yts Does the proposed change: 1. l"ake changes in the facility? ~ 2. !".ake changes in prececures as they are described in the F5AR7 3. Involve test cr-experiments not describe in the F5AR7 / 4 Require a change to the technical specifications? ~ / Any ansver to questiens 1 thrcugh 3 "YE5". then a 10CFR50.53 evaluation is required. FEAR Change Recuest per precedure SN404.01.C0 may also be recuired. Ancuer 4 is 'YES". then Technical 5pecificatien Change Request per precedure $U404.01.C0 and NRC approval is required prict to implementation. All answers 1 througn 4 are *!D". no 107R50.59 Evaluation required reccernend acticn a;;:reval. ~ 10CFE50.59 EVALUATION (Previde Response Jus *.ification uith References) l i 5. Will the pretacility of an accident previously evaluated in the F5AR be increasec7 1
- 6. Will the censesquences of an accident previcusly evaluated in the F5AR be increasee?
l 7. Will the p;ctatility of a malfunction cf equipment impcrtant to safety te increased? 1 l B. Will the consequences cf a malfunction cf equipment impcrtant to safety be increasec7 / 1 S. Will the possibility cf an a=ident cf a different type than any previcusly evaluated in the F5AR be created? y
- 10. Will the possibility cf a malfunctioning of a different type than any previcusly j
e ) evaluated in the FSAR te c;tated? p ~
- 11. Will the margin of safety es defined in the basis for any technical r:ecificaticn be reducec7 l
Any answer to e,uestions 5 thrcuq511 "YEE". then a pctential unrevic ed safety cuestien is identified. Pr= red to prececure 7 x07.C3.00 prict to iro!r entaticn. j j All answers 5 thrcu;n 11 are ";;0". roccerend ccticn e;preval. If F5Aff che:ter 5/Chacter 15 is ;ctentially ef fected, fer.ar:: a c=y cf evalus:icn to 1.uclear Fuels hina;r ent. I verify that the abcve reviedevaluation is acewate and accurate are that the voertief.ed have received rem i:cd trair.ing. n W ! I. [ h 4=c.l a=k $ ~' 46 mhn r Wa /.);;;;AX; [ p*E / I'...: ~; r i :;ci: ni' A i.i. C.;" A ~
i l l <.. 1 107P30.59 REVIEW AND EVALUATION l REERNEE JUSTIFICATION race _I_ cr 3 Acilen UNOER REVIEW: Mc' A/ as , R1 REvis o, 1 Nema / Ti t le FRCCEOURE/PCP/TDP H00. NO: / t l CUESTICN l RESPONSE JUSTIFICATION E lNA M M d w clo ov>f 1sudve m l b =j % NMkh &/or k%k-J <rda dh d AcoA d of' A d lh *&g)M as -btalt %%wi a y, r L > b % b a b C h i 6 ) a s u achQdJ. l Irrev?GsaAs Qvahuttsd" 't4 k FW w a e( O be IWta-ad b,y M i & ' ' Onvusum o >e 6m a lc.ik} ~1 atnnuol u ,' CVO Y b (b f$AQ (,nj Q & Q { ' ' Ir m J2 M d b e m 7Jr.e, g ge g 4 , f L FMR arit e n ;ag ja g l @ *% ds ord shrvolve m y* %cemm u em&i,4% p a l'% og l G^) m * @ @ w % f l a ve he w w % g ,mm,wws ~ w 7,l % dumwn do or.J rkcujt is n g l S*hr, ;%wfvo d c67,h s w g g g 3 4 'doso k !(,as;6c 9) e u I / fu
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ICCTR50.59 REVIEW AND EVALUATION RESKrNSE EST!FICATION PM CF ACTION UNDER REVIEV: ggyg5 cy f Hame/IItle RECURE/PCP/TDP MOD. N0: /\\[ (( l / i STICN l RESPONSE EST!FICATION I ~lh h>,h al W i} & ah'n L 2 y M WMbk or of yeomy ;%c 'N i L T*d w& % Auk, nu' l M ~n n d le tJ u Leg u %A B M%g %g m
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.2[IS~ . CNSITE POWEA s7 STEMS 'G-I the unii auxiliary tr.:.cfornzr, and the Class IE buses normally a r e s q. :..'..,. th o t ;h :.. t s;.rtup transformers. In the e~ vent of-failure of-the unit auniliary transformer, a turbine. trip, or - reac c:. trip, an autcmatic fast transfer of the 13.8-kV buses to the startup 't' transformers is initiated to' provide" power ~to the au: ciliary,lodds - Transfers'of 'all buses '(subjedt 'to limita2 e tions of power system development, sec.ticn 8.2.1.2.1) can be. ~ initiated by the coerator from the control room. Figures 8'.3-1 and 8.3-2 shew connections of powefsupply feeders and busing arra.ngements of the ac power sys. tem and list loads supplied from each bus.,; Preferred pcwer for Class IE bus,es is supplied. from the startup _ transformers through the 13.8 kV switchgear and the,13.8 to 4.16 kV ESF ' transfor.mers... ~ ~ Reactor' Edoldnt p6mps'1A and 2A are~ connected to one'13.'8 kV r P bus 'an'd 13 an'd 23 are cor5ected to'the other 13.8 kV bus. ~ / Electrical supply for~rea~ctor coolant ~ pumps is arranged so l"*- that the pumps will normally Femain. electrically connected to J
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...( l the' turbine generator for 20 t.o 30 se.co_n.ds following a turbine trip should offsite power not.be a.va.i.lable.. Cr. edit is not taken for this. feature.
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....e .s, 8.3.1.1.2. Non..C..las.s IE Ec.uipm.ent Cana iti. es... good.4 f.'- l13.8 kV Suitchgen:
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. 3000A 1nuous Rating 1 '- n 5Q6,b,YiVW)br... n.- MAN-SO2 acing. 2.. . NAN-503 ^- 4 pA3 sc4 c. NAN-SOS ^" NAN-506 C ~ . "Inc55ing 'sre a:fers '2000A Continucus, 40.2 kA interrupting at 13.3 kV voltage Feeder Breaker: 1200A COntinucus, 40.2 kA ' interrupting at 13.3 kV /oltage e l Aten._-:.. 3 5.2-2 August 1932
T,703 FSAR' sa6a M,$. / ONSI"'E POWER SYS"" EMS .., ) 8. 4.15 kV Svitch p12 7 ooo A O.: c ; ~. '1. ~ - 1.". . itinuous rating - t 3000A C NIM-SC2 ,/fpfA bracing ^ Incoming and' Tie 3000A Continuous ~ rating - 47 kA Breakers , interrupting at 4.16 kV volttga Feeder Breakers 1200A Continuous rating - 47 kA interrupting rating at 4.16 kV voltage C'. - 480 Volt Unit Lead Centers Transforr.ers 1500 kVA 13.8 kV/480V,'3 phase, 60 E: 1000 kVA 13.8 kV/480V, 3 phase, 60 H: 300 kVA 13.8 kV/480V, 3 phase, 60 2: s ,.. ) Buses f. 3000A Continuous (1500-kVA rating) ~ 1600A Continuous (1000'kVA rating) 600A Continuous (300 kVA rating) s. ~ r, Breakers (Metal Clad). . ~.. - 600A (non-fused) 30. kA ' interrupting rating at .480V.. s 600A (fused) 200 kA interrupting rating at 480V. 1600A 50 kA interrupting: rating"(at 480V). ~: 2000A 65 kA interrupting rating at 480V. 2000A 65 kA interrupting rating at 480V D. 4BOV Motor Centrol Centerc ~ Ecri ntal bus 600A Continuous Vernical bus ' 300A' Continuous D. August 133; 8.3-3 Amend.cnt 9
AJ 6 2. PVNGS-FSAR %5 ONSITE POWER SYSTEMS Breakers (100?.t:/2 100/IFc JA 225?. Frame) ~ .,, q \\Qoo0/;5;oco/ r,om425000A (Magnetic) j; G 2 00 0?. 'Th =:2 M:p tic) 25000A (Thermal Magnetic) 225?. ~; = :nly) Breakers (400?. to 600A) 30000A (Magnetic) 20000? 'The_m l M:p atic) 8.3.1.1.3 Class IE AC System The Class IE ac system is that part of the onsite power system inside the dotted-line enclosures shown in figures 8.3-1 and 8.3-2. The Class IE ac system distributes power at 4.16 kV, 480V, and 120V to all Class IE loads. Also, the Class IE ac system supplies power to certain selected loads that are not directly k safety-related but are important to the plant. Table 8.3-1 lists the safety-related loads supplied from the Class IE ac system. The Class IE ac system contains standby power sources that automatically provide the power required for safe shutdown in the event of loss of the class IE bus voltage. i Voltage levels at the safety related buses are optimized for the full load and minimum load conditions that are expected throughout the anticipated range of voltage variations of the power source by the adjustments of the voltage tap settings on the transformers. The following describes various features of the class IE systems. 8.3.1.1.3.1 Power Supply Feeders. Each 4.16-kV load group is supplied by two preferred power sypply feeders and one diesel 8.3-4 1
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1 l pm ~ PVNGS FSAR ONSITE POWER SYSTEMS Each 480V load center' bus is equipped with an under-voltage relay for annunciation "in the c'o' trol room. n ~ ; ;~ ' ' _~ ~s. C. Differential Relaying ~~ .... ;. j v.
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?........ ~ Circulating water. pumps, react..o..r..c.oolant pump motors, diesel generators, and transformers larger than 10 MVA u.,..- .v ,... e., s, s m _, are equipped with ' differential Yelars..These relays ,......,..eu,..-- .:..?,.. .... provide high speed.d; disconnection to u.,, prevent severe ...,,,.m.. g... damage in case of internal circult" faults. . w : 2; :. :c- .480VLoadCenterProk<. tion D. m
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.:. ::. :. % 3. - Each load center main feeder circdit is protected by three phase relays and one grounc1 overcurrent relay. Each motor control center main feeder circuit is pro-tected by a circuit breaker equipped with an adjust-able, selective overcurrent device in each phase and by one ground overcurrent relay. Motor protection is provided by three phase relays which alarm on overload and one ground overcurrent relay that trips the breaker on ground faults. Trip will occur only on faults or sustained overcurrents substantia 11y y eater than full
- a. ggto safe shutdca:n c4 ihc. PM load current.
e E. 480V Motor Control Center Relaying Molded-case circuit breakers provide.. time overcurrent { and/orinstabtaneousshortcircuitprotectionforall connected loads /. For motor circuits, the colded-case circuit' breakers are equipped with instantaneous trip 'only. Motor overload protection is provided by heater-element trip units in the motor starter. The colded case breakers for nonmotor feeder circuits provide l thernal ti=e overcurrent protection as well as instan-taneous short circuit protection. Thermal overload _j protection ofh safety-relatQsLif automatically _ hvnamed fellowinc an ESD.Sr T-hermaL_overacac_pmD r 7.] ' " n--.f-or-motor %per,a ted--ve/es-below-1-HP-i nside-the-i C e l -- N w fdcepc" Sprcn4,bCrd beccIEr i. C Crd h u r.Yle-, N% W >t l ' e' CL m
4A6A lhent:o.l oveelc[. dbc. bon. kir5dE5Ec[cd Wbes, bboe l HP' 71 t 'Mide vc.cnicubrre % Mh d ISO Percenk Of. Mmed A.,cf p a d C.jh2- ^ I
-- -onsiTE POWER SYSTEM
( M containment in-bypasseo--twnsurc adaquate fault luvel[ ' tim char.ac ter4+t-ics-for-proper--b e* M-+4 T (theclectricalpenetratienc-. ,,The bypass circuitry is designed to IEEE 279-l.971.cr.it.eria...n .,_s The short circuit prote'etive system 'is " analyzed to"en'dur'e 'Ulat - the.various ' adjustable devices are applied 'witihin"t[ heir' rat'ings and are set to be coordinated with each othe'r to' 'sttain' ' se'le tivity in' their operation. IThe "combirIatiori 'of devices and " - . settings applied affords the " selectivity nece's'sary 'to isolat'e' a faulted area quickly with a'minisum'o'f'disturban~c'e'to the O ' rest'of the system. l r W"r " ~ N R 7' " ~4 ~ Preoperational' test of the p'rotective'n v.: devices.are performe.d to demonstrate I. hat they are properly calibrated and adjusted to a alarn or trip as required. 'After the plant is in operation, periodic. tests are performed at scheduled intervals to verify the protective device calibration, setpoints, and correct ( operation. A random. sample of molded case circuit breakers required for the protection of containment electrical penetra-tions will be tested periodically to verify the overcurrent trip setpoint in accordance with IEEE-20. The PVNGS design has a combination of fuse and circuit breaker, two fuses, or two circuit breakers for circuits being fed through containment electrical penetrations. 8.3.1.1.3.14 Testine of the AC Systems durina Power 0:eration. During periodic class II system tests,.,subsyste== of the i Class IE system such as safety injection, containment spray, and containment isolation are actuated,"th'efehy causing,.. ,s appropriate '; circuit breaker 'or' c'ontactor 'o;IerEtion. .The'~ 4.16-kV switchgear and 4807 load center circuit breakers ab.d control circuits alro can be tested iridependently while individual equipment is shut down. The circuit breakers can be placed in the test position and exercised without operation n' of the associated equipment. 9
e GG6G PVNGS FSAR .1/ ' o /5 ONSITE POWER SYSTEMS ~ ~ - 8.3.1.1.4.1 Automatic Starting Initiation Circuits. diesel generator is automatically started on any of the fol Each . lowing conditions. @housta2d'degradah' Undervoltage on the 4.16 kV Class IE bus to which the e generator is connected, loss 4cf offsite power (LOP) . Safety ~ injection actuation signal (SIAS) o Auxiliary feedwater actuation signal (AFAS) e 8.3.1.1.4.2 Diesel Generator Startina Mechanism and System. The diesel generator starting system is described in section 9.5.6. The design basis and analysis for diesel generator -systems controls and instrumentation are described in ' tion 7.3. sec-8.3.1.1.4.3 TriDoine Devices. Devices are provided for the following protective functions for each diesel generator: Incomplete sequence (start failure) o Engine overspeed e High jacket coolant temperature e High bearing temperature e High crankcase pressure e Excessive engine vibration e Turbo-charger thrus t bearing failure e Low lube oil pressure e Turbo-charger low lube oil pressure e e Loss of field 'e Generator differential Generator ground overcurrent e 8.3-19
[ l.. Dab PVNGS FSAR ONSITE POWER SYSTEMS B. Interrupted auxiliary feedvater flow to the steam' generator (s) is fully reestablished within 23 seconds. The deviation from the CESSAR requirement C-14 of 15 seconds is acceptable to Combustion Engineering as discussed in section 1.9.2.4.10. 8.3.1.1.4.7 Instability. 15 Refer to section 14.2, and Section 3/4.8 of the Technical Specifications for testing requirements. 3l During testing it an SIAS or AFAS occurs while the diesel generator is paralleled to the pr 5-erre 3l control suitelr-T5'~the-REMOT over supp1Ntbthe or LOCAL position diesel N generator breaker will tre automatica11y trip
- ed, b p
f a r. v_ a,_, f tri;;ir.; pu1 W. The diesel generator will,' continue running and automatically-revertPhe isochronous rotectiv jeevices are bypa \\ oder All non-critical [ / N ssec. I a non-criti'calNip-occ./ 3 during testing, the diesel generator will trip. urs On a subsequent \\ 14, SIAS or AFAS or LOP the diesel generator will automatically start and run in the isochronous mode. The LOCAL control position 'is selected from the local control panel for diesel fenorator maintenance tasting. A diesel generator LOCAL POSIT 1CN alarr will be annunciated in the control room. To prevent any starting of the diesel generator during maintenance, the 07? position is selected at the local control panel and.,ar DIESEL GFEEEATOR INOPERABLE alarm is initiated at safetQ5iuipmentstatussystemannunciator. the l If the protirtad power source is lost while paralleled to the diesel generator during testing. the diesel generator will trip on overcurrent and the diesel generator breaker vill trip automatically on a diesel generator shutdown signal. Upon detection of undervoltage on the Class IE 4.16 kV bus. load shedding and sequencing vill be initiated as described in section B.3.1.1.4.6. \\ Amendment 15 8.3-44 April 19a4
Bobi /of PVNGS FSAR ONSITE POWER SYSTEMS ~ " ' " Table 8.3-4 120V AC VITAL POWER SYSTEM LOADS (Sheet 3 of 3) Channel D (Cont.) Plant Protection System (PPS) Main Control Room Control Board 12SV DC Control Center E-PKD-M44 Scace Heater MOV Position Indicators at 125V DC Control Center 10 Inverter Space Heater There is no prevision for automatic loading or load shedding of the buses. Inverter trouble and bus undervoltage is annunciated in the control room. In addition to the 4 inverter power supplies, two additional 480 volt, 30 inverters from channel C and D batteries, supply dedicated powcr to the shutdown cooling motor-operated valves. 8.3.1.1.7 Non-Vital AC Instrumentation and Control Power Supply The 12CV Non-Vital ac instrumentation and control power supply furnishes power to Non-Class IE instrumentation and controls. The Non-Vital ac instrumentation and control power supply for each unit consists of four tr 3t rrptdtri yr ste r.c fo rmere (two grounded and two ungrounded) and 120V distribution panels as shcwn on fig-ure 8.3-4. Each distribution canel is normally sunelied u t teqe Y4 :, des through the associated t-r-ansfoher frca a non-Class IE MCC Each 120V distribution panel has the capability of being sup-plied by the dicscl. Ecwer tc ($hh$1scurc+ is tricced on an SIAS signal and can be re-established manually after the sequential loading of the dicscl generater._ % e, two (CL, & NftmbrfneCS ~ ^,c6N tJ t fjlAlo.}015 l l L_______---___---_------- i
Mb: PVNGS FSAR uph-APPENDIX 8A (. circuits or common mode failures in the sequencer design that could render both on-site and off-site power sources unavailable. Furthermore, we would require the applicant to provide the following additional information: 1. A full description of the load sequencer design feature in the FSAR.. This should include sequencer power supplies, test features and alarms. 2. A reliability study on the sequencer.
RESPONSE
A sequencer design demonstration test was per-formed to test the sequencer to assure that no credible sneak circuits or common mode failures could render both on-site and off-site power sources unavailable. The testing included approximately 130 credible scenarios combining accident situations with and without off-site power avail-able. The test results were satisfactory and demonstrated that no sneak logic paths exist in the design that could result in failure of the sequencer to perform its required function. The test report "ESF Load Sequencer Design Demonstration Test Report," E160972, February 1981, which was submitted to the NRC. Item 1: ouSEdbe2 EG9fd @ f A. Load Sequencer Design 3 Each redundant ESF load sequencer system cerforms logic functions to generate the losslof off-site power (LOP) signal / load shed signal, the diesel generator start signal (DGSS), and the load sequencer start and permis-sive signals. Each redundant ESF load sequencer system is supplied from a separate 120 V vital ac distribution bus and a separate Class 1E 125 V da distributim. cus. October 1981 9A-21 Arendment 6 . _ - _ _ _ _ - - - _ ~
I + e t A.26 p PVNGS ?SAR 'S(3- =n^ (ttdo6d8N_ W APPENDIX 8A The LOP signal / load sted signal logic continuously monitors the Class lE 4.16 kV bus for an undervoltage condition using -- undervoltage relays. If an under-voltage trip occurs, annunciation and indication is provided to the operator. On a 2-out-of-4 coincidence of undervoltage relay trips or upon manual actuation, an LOP signal and load shed pulse are generated. The LOP signal is stat to the DGSS logic. The LOP signal. (r.intained through a 60-second off delay) also actuates .% ced shutdown system loads by de-energizing actuation relays. The load shed pulse (1 second) sheds 4.16 kV anf ;clect;d I;^ v loads from the Class 1E 4.16 kV bus and trips the 4.16-kV Class lE bus preferred (off-site) power supply breakers by energizing actuation relays. The DGSS logic combines the LOP, SIAS, AFAS, and manual actuation in a logical "OR" to generate a DGSS to start the diesel generator. 6 The load sequencer start and permissive signal logic monitors input signals, determines the appropriate mode of operation, and generates sequentially timed start and permissive signals to ESF and forced shutdown loads as required to prevent instability of Class lE buses. Start signals actuate devices by de-energizing actuation relays. The permissive signals, however, allow loading of devices by energizing actuation relays. The load sequencer controls only pumps, fans, and chillers, and does not control any valves or dampers. As such, the load sequencer does not cause complete ESr system actuation. The load sequencer responds to the following conditions: LOCA, with or without off-site power available o Accident other than LCCA. with or without e off-site power available l l 1 1 6l Amendment 6 8 A-E ?"
I ____. __ - _ - \\- \\
- pasos PVNGS FSAR
/.Eh / ONSITE POWER SYSTEMS 7: The consequences of frequency decays of up to 3 Hz/sec (with bus voltage at its nominal value and with all RCPs connected to their buses) on the Reactor Coolant Pump buses are not more severe than the consequences of loss of flow of the four RCPs due to loss of power. The Applicant's RCP buses, therefore, shall not subject , the RCPs to sustained frequency decays of greater than 3 B:/sec. 8. The following tables provides electrical data for the safety-related equipment which is generally supplied by Combustion Engineering. Complete tables and responsi-bilities for supply will be provided in the Applicant's Safety Analysis Report. Table 8.3.1-1 Power Requirements for CESSAR Design a. Scope Safety-Related Equipment. ( b. Table 8.3.1-2 Pcwer Requirements for CESSAR Design Scope Safety-Related Equipment at Various Operat-ing Conditions. Table 8.3.1-3 Power Requirements for CESSAR Design c. Scope Safety-Related Electrical Equipment During Emergency Operation. \\ d.s Table 8.3.1-4 Required Standby Generator Loads T 9. The vital instn: ment buses shall be designed such that N 1 the maximum voltage fault shall not exceed 480 V;sC + g 10kor 5 VDC + 10%. ~ 10. Cabling shall meet the requirements specified in Sec-tions 7.1.3, 7.2.3, and 7.3.3 for separation and inde-pendence so that no credible fault in ene channel can \\ be propagated. \\n - i l
l I Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The DAVPS l l facility is an independent structure, Ismotely located, with no interaction with safety related equipment. In addition, the functions carried out at the DhWPS facility are not required for the safe shutdown of the PVNGS unit. The change does affect an important to safety system, the fire protection (FP) system. The FP vater for the DAWPS facility is supplied by a branch from the main FP piping for the l plant and all FP piping has been designed in accordance with all applicable NFPA requirements. This modification has also been evaluated for compliance with the applicable NRC regulations in accordance with IE Circular 80-18, "10CFR50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems " and Generic Letter 81-38, " Storage of Low-Level Radioactive Waste at Power Reactor Sites". i (43) Description This change revises FSAR Table 6.2.5-1 to reflect the as-built design parameters of the containment hydrogen control system. The correct hydrogen analyzer scale is 0-10% hydrogen, and the correct accuracy is 3.0% full scale. In addition, FSAR Table 7.5-1 was modified to reflect as-built system capabilities and to be consistent with FSAR Table 6.2.5-1. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safetv Evaluation This change did not introduce an unreviewed safety question. The hydrogen analyter retains its safety related seismic category I design to preclude failure of the analyzer function or other important to safety equipment. The analyzer is redundant so that a single failure will not remove hydrogen indication. The point at which the recombiners are actuated (3.5%) falls in the 0-10% scale of the analyzer. )( (44) Descriorien This change revises FSAR Section 1.8 to correct the design class.for the primary safety relief valve position indicator from Q9E to R9E and the power supply from 1E to non-1E. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safetv Evaluation This change did not introduce an unreviewed safety question. The change is consistent with the guidance of Reg. Guide 1.97, Rev. 2 and NUREG-0737, II.D.3. The function of the indicator is not changed, which is to provide post accident monitoring. - _ _ _ _ _ _ _ -
p j!', '. ~,, 297 fg FILE: W[- 03 0-Y4'/ PAGE / OF 4 REVIEW AND EVALUATION ACTION UNDER REVIEW F 9 R~A b p u b d OM e / 8 REVISION: l DESCRIPTION OF PROPOSED CHANGE: bn (A D d >vJn43 0Ono A a d % dan w s.< o n,n 01 A2 M h i A$4 VJ c!M bh a, W) s 6 1 VMn amAw%q,m tia_ _ 7 v y oq g 10CFR50.59 REVIEW DOES THE PROPOSED CHANGE: 1. MAKE CHANGES IN THE FACILITY? YES / NO 2. MAKE CHANGES IN PROCEDURES AS THEY ARE DESCRIBED IN THE FSAR7 YES NO / '3. INVOLVE TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR? YES NO / 4 INVOLVE ANY OTHER CHALLENGES TO NUCLEAR SAFETY FOR PVNGS? YES N0 - 5. REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS? YES NO / 10CFR50.59 EVALUATION (Provide Response Justification with References) 6. WILL THE PROBABILITY OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES NO / THE FSAR BE INCREASED 7 7. WILL THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES NO2 THE FSAR BE. INCREASED? 8. WILL THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES NO / TO SAFETY BE INCREASED? 9. WILL THE CONSEQUENCES.OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES NO / i TO SAFETY BE INCREASED?
- 10. WILL THE POSSIBILITY OF AN ACCIDENT OF A DIFFERENT TYPE THAN YES NO /
ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 11. WILL THE POSSIBILITY OF A MALFUNCTIONING OF A DIFFERENT TYPE YES NO /
THAN ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 12. WILL THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY YES NO /
TECHNICAL SPECIFICATION BE REDUCED? ANY ANSWER TO QUESTIONS 1 THROUGH 4 "YES", THEN A 10CFR50.59 EVALUATION IS REQUIRED. FSAR CHANGE REQUEST PER PROCEDURE 5N404.01.00 MAY ALSO BE REQUIRED.- i ANSWER 5 IS "YES", THEN TECHNICAL SPECIFICATION CHANGE REQUEST PER PROCEDURE 5N404.01.00 AND NRC APPROVAL REQUIRT.D PRIOR TO IMPLEMENTATIO ANY ANSWER TO QUESTIONS 6 THROUGH 12 "YES" THEN AN UNREVIEWED SAFETY QUESTION IS IDENTIFIED. PROCEED TO PROCEDURE 7N407.03.00 PRIOR TO IMPLEMENTATION. 1 ALL ANSWERS 6 THRU 12 ARE "NO" RECO'D!END ACTION APPROVAL. ALL ANSWERS 1 THROUGH 5 ARE "NO", NO 10CFR50.59 EVALUATION REQUIRED, RECOMMEND ACTION APPROVAL. I verify that the above review / evaluation is adequate and accurate and that at least one of the underni:;ned has receis ed required training. N Oh (2p 31 ) !) lD $ .Y A fl% k /% 0 ) mam ,m a m uoR m m om,
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4. &L413 6[y 7 10CFR50.59 REVIEW AND EVALUAT ION RESFONSE JUSTIFICATION PAGE OF d!> ~ ACTION UNDER REVIEW: IU4bk ^ bOM On *ht bBp f, k / REYlSION: , Q Name/Ti tle FROCEDURE/PCP/ TEMP MOD. NO: 1 CUESTION l RESPONSE JUSTIFICATION 0 W_Y00110A)ADDD ll/Y\\ 0 l\\h t h fn M II It w n J OQrn wINa ,h 1 Rn 0% ln i cr1 /%rD lan) unoca enn k n P o w ifvno,m u IkO0 tJJA o/nn Am Y, OA D}M rl o A W1 A.A 3 (YY) 0 ('tsyt}) I1hddsid,a l-Ju On-)MwnYi4 KlY%oNNica bN.$c)m kb W0n sD n~ Ann,wx svmfi, lJ b llDnh 0 0 A D Q 3hsanoA .M (YY)O J IJ)'li / ) #$ l 0 0 9 0 41_hD /,*YYl A />A O- ^ Arw kO <:s 11!YY Ay1YO I A / Y.OD fh? -}AOO 0/0.*YY)1h () [nlu.+r h u w ro V U _0 WkSQt i J% 0 OUY fYO Q 0!J l/HA1 bf10 l1M 0 0 m n t.0 LM D/ O /2 A O OP fl a A D W RO l !) AY.)J O LM 10$6 neo n ),x 9 or k inn nOns a,,A N,e,,v<>, do)n ~ . Y Oh _t l f,0, k.! r s o O r,. e. M Y D ) !. i ek n k 02.P t r) ? - Ikh ' 1 (YW2 c~~ ...n i f,4D Dn O Y oA ll.l 0 inV', A;No X2 trna!%indikn n ~ O O n, &J )w l /1on+ Lt D n,D n Nr. / 5,_.l71e 1/3 r. i /2l$ enkw aw] % dL /Ix9 \\bl:00 ki-o m,,) ~ i ~ u (s ,,) Q _. j7m n , h / i s' Y bh- .lHA / b 7/ kv le we'r Cot e \\bvlever Supv' or SS Da' t e / ~ Form 7N407.0tB Rev. l L__ _-
7 ^* fg f}. ~ 10CFR50.59 REYlEW AND EVALUATION RESPONSE JUSTIF1 CAT 10N P' AGE OF M ACT10N UNDER REYIEW: MhI /T) /) n & dB,nij M /7 ~ G Hame/Ti tle ' ~ REY 15 TON: ' FROCEDURE/PCP/ TEMP MOD. NO: .I _00ESTION l RESPONSE JUSTIFICATION OA 1Y WJ \\$) ATOA C A\\ b(O l,,,g,%,) G O O (I ' '0 /Y\\ fn e i I I i l i I i i l w" I l 1 I 8-1 I I I I I I 1 l l l 8 l 1 I 1 l 3 $ '? ! )AL 0/tv W j Dihp 0' $0 [ " Pavlever Sy6v er SS '+ Reviewer Cafe C6t e / Form 7N407.o:B Rev. l
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e .4 6 (55) Description The description of the Train A safe shutdown raceways for fire zone 10B in FSAR Section 9B.2.1.1.C.6 were changed from a 3 hour rated wrapping to a 1 hour rated wrapping. In addition, the fire severity times were corrected accordingly. This change was transmitted to the NRC in the USAR, Rev. 0.- Summary of Safety Evaluation This change did not introduce an unreviewed safety question. Zone 10B has both fire detectors and an automatic fire suppression system, therefore, only a 1 hour rated wrapping is required per 10CFR50, Appendix R, Section III.G.2. (56) Description FSAR Section 12.5.2.2.5 was revised to define the calibration frequency for flow measurement devices used in conjunction with portable air sampling instruments and equipment. This information was previously not discussed in the FSAR. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. This change provides information not previously discussed in the FSAR. The portable air sampling equipment has no impact on accident analyses or equipment important to safety. (57) Descrintien The FSAR Section 13.1.1.2.3 description of the activities of the Nuclear Construction group was updated to reflect the responsibilities of the group once the construction phase is complete. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Saferv Evaluation This change did not introduce an unreviewed safety question. .The change will not affect the ability of the plant to operate safely or the operation of equipment important to safety. Safety related activities still fall under the QA programs, and construction will be done with appropriate procedures. [ (58) Description FSAR Sections 9B.2.12 and 9B.2.15 were revised to add deviations for fire zones 73, 37B, 37D,
- 39B, 74A, and 74B to reficet the actual thermolag installation in those fire zones. The fire barriers are in l.
untested configurations. This change was transmitted to the ::RC in the USAR, Rev. O. l l 18-f l 1 L-_ _ __ _ ____------_____ -- J
.i 'S Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. This change allows a deviation to 10CFR50, Appendix R, Section III.G.2 for -the configuration; however, the configuration provides the equivalent protection required by III.G.2. (59) Description FSAR Section 13.1.1.2.2.2 was revised to delete emergency planning from the responsibilities of the Technical Services Group. The Assistant Vice President, Nuclear Production is responsible for emergency planning, and the removal of the responsibility from Technical Services was overlooked when the responsibility description for the Assistant Vice President, Nuclear Production was changed in Amendment 17. This change was transmitted to the NRC in the USAR, Rev. O. Symmarv of Safety Evaluation This change did not introduce an unreviewed safety question. The emergency planning group still exists and their function has not changed. The reporting was changed to increase the attention given to the group, and this change was administrative, to correctly reflect the reporting chain. (60) ' Description FSAR Section 13.1.1.2.2.3 was revised to reflect the following organizational changes within the Nuclear Engineering Department. Onsite liaison was renamed Resident Engineering. Operations Support and Scheduling was deleted. The Projects Supervisor position and the Methods, Training and Compliance group were added to the department. In addition, the Resident Engineering and Projects Supervisor report to the Nuclear Engineering Production Manager. This change was transmitted to the NRC in the USAR, Rev. O. i Summarv of Safety Evnluation This change did not introduce an unreviewed safety question. The Nuclear Engineering Department will continue to provide the same
- support, fulfill the same responsibilities, and perform the same function as before. The change was made to enhance the performance of the department.
(61) Description FSAR Section 13.2.1.3.1 was revised to replace the requirement for " lectures" in the modified systems course for non-licensed operators to " training". This change was transmitted to the NRC in the USAR, Rev. O. l l I 1 i
Jd6 s l 5~~ 4 FILE: R7_03,J </ O'/# PAGE _ l OF ' 10CFR50.59 REVIEW AND EVALUATION ACTION UNDER REVIEW: hrVAw qu kr Postcm3 (mWLm REVISION: O N ti#.K $E4. %.L.it 4 55,L. it' DESCRIPTION OF PROPOSED CHANCE: btw e eas rest m wt s % l% Wy 'l6,51D S 3# 6 W LeccPr hhoturm o ok bub. kn howw twD \\ ewet m \\ __ 'P/cvion AnRerrJa T&o revne,a 10CFR50.59 REVIDJ DOES THE PROPOSED CHANGE: 1. MAKE CHANGES IN THE FACILITY? MAKE CHANGES IN PROCEDURES AS THEY ARE DESCRIBED IN THE FSAR YES / NO 2. YES NO v/ ( 3. INVOLVE TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR? 4 INVOLVE ANY OTHER CHALLENGES TO NUCLEAR SAFETY FOR PVNGS? YES_ NO / YES NO v 5. REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS? YES NO / 10CFR50.59 EVALUATION'(Provide Response Justification with References) 6. WILL THE PROBABILITY CF AN ACCIDENT PREVIOUSLY EVALUATED IN THE FSAR BE INCREASED? YES N0_ / 7. WILL THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED THE FSAR BE INCREASED? YES N0 g 8. WILL THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES N0_ i - TO SAFETY BE INCREASED? 9. WILL THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTA YES N0 TO SAFETY BE INCREASED? y 10. WILL THE POSSIBILITY OF AN ACCIDENT OF A DIFFERENT TYPE THAN YES. NO v ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED? 11. WILL THE POSSIBILITY OF A MALFUNCTIONING OF A DIFFERENT TYPE YES NO V THAN ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED? 12. WILL THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY YES N0_ / TECHNICAL SPECIFICATION BE REDUCED? X ANY ANSWER TO QUESTIONS 1 THROUGH 4 "YES", THEN A 10CFR50.59 EVALUAT IS REQUIRED. FSAR CHANGE REQUEST PER PROCEDURE 5N404.01.00 MAY ALSO BE REQUIRED. ANSWER 5 IS "YES", THEN TECHNICAL SPECIFICATI0" CHANGE REQUEST PER PROCEDURE 5N404.01.00 AND NRC APPROVAL REQUIF.D PRIOR TO IMPLEMEN l ANY ANSWER TO QUESTIONS 6 THROUCH 12 "YES" THEN AN UNR QUESTION IS IDENTIFIED. PROCEED TO PRCCEDURE 7N407.03.00 PP,IOR TO l l IMPLEMENTATION. 1 ALL ANSWERS 6 THRU 12 ARE "NO" RECOS!END ACTION APPROVAL i' ALL ANSWERS 1 THROUGH 5 ARE "NO", NO ICCFR50.59 EVALUATIO" REQUIRED l RECO.W END ACTION APPROVAL. I verify that the above review / evaluation' 1 is adequate and accurate and that at least one of the undersi:;ned has received required training.. 5_ ffk /'9k?~07 e m[i m h ^ / om mmm.,,mm om
c7d87 iw 13/l9_ l l 10CFP50.59 PEVIEW AND EVALUATION RESPONSE JUSTIFICATION 1 ACTION UNDER REVIEW: F Vt 4Tichi TO hPf_ Fam:ImW 6nbfufL REVI SION: O Name/ Title i FRO'EDURE/PCP/TEPP M00. NO: b b l 1 CUESTION I RESPONSE JUSTIFICATION I k I \\M C\\mKWG. 1AMES 15 C.64sOCW ld Tpy_ i:cAtt t N TO I i ?i4LE.CT TMf trTUA mt0RIAoha it KTn,LL(hTMO M 80 PE 7<d % t i ,M h.alb b bsM b, lib. l 3 1 blMCC. 'Th5 CAWMcE IS d6 \\d hN 9/LOCEt1%EE, TBL ?PALEDOEAS I bc DESCRio+6 to m W F % CL b?c. Merr hpMcwA5 I b 13ts filmO66-DCES VK3T" b30CbL rE t,TS O P_ Extyttu p prs ) jot l I hE ECRefb 'o w FEAR mum m ws ss
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~ hOQ 'g 10CFR50.59 REVIEW AND EVALUATION RESPONSE JUSTIFICATION PAGE b OF ACTION UNDER REYlEW: Ed AttM "ID E-E. (o M SDd 4/lTE/_44 REYlSION: C) flame / Title PRCEECURE/PCP/TEtP M00. NO: 1 OUEST10N l RESPONSE JUSTIFICATION l ) 11 (Mt35E-QOf43tf 5 OF Ad A004DE M EMMCASLV 6@UMVb td i i m. Fm2 t.c. O ibr ise-etaCarnsch eerSosc Tue. Pua e i I OIvrbootAmtw) P P_bth D M EQN9skEnv 9/crm CThi Ac, 76001d#43 I Y Ih $$O)OOO s I d 7MABiLITY CP 4 M AlF1tatmBS of F60tPREcr #o 0o02%?,r I i *TD 54F1E N LOtLA-13CW EE-IdC(%YJA3 'dECM56 00 C+hV 36 CE-S 3 I% Et90!PsAf.Mr Mf46 MhbE-. I tM
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S. 3681 tf/,,. 100rR50.59 REVIEW AND EVALUATION RESPONSE JUSTIFICATION PAGE_ 7 0F ACTICN UNCER REVIEW: eejiATt m W h E4,. ecycyiod [19mn gg ggyg53cn; g Name/ Title PRCCEDUNE/PCP/TEFP MOD. NO: 1 CUESTION I RE'aPONSE JUSTIFICATION I I D tBlu-N E A LA Ny_nodp S ok l( bg(ft%y 1 I W PC TMG AOV 9,tEd&J5LY Ev%tll b T h (Y "TME-b Ik k lI (db Ocr 66-O f-D1Eh W GNFS Ttt01?E. MS-VM 4Whiant i I (Y D65.Il-d (.NOdbES. f5 (4mAVM ALL.04d5 h Dt/l A~i'lOM i l Tb R nod W.b.Z h EQUnn!ec Pr7pTCT103, 17.- flu IA M7iJ oh SAFEW AS DE AtJEb 93 "r M Leis Feft i l 60Y TECR4 h0L Ef6 CI F icATied uhtu e)c r EE. REhXL.h 1 I.51QCE 9 55 t % ( G Xl f., 15 iDT DE EC.#AEB I4 M I l ~~TEf;H OUk L. SPEO Fi CAT \\tw%, 1 I I I i l i I i l I I I l l l l l 1 /4-3rST '\\ r'-- / * '" ' I __ lp,t ? ,y r-s. 3 .,negg, bhTE. beioe,1 h v L+*i E l (' Form 7N407.010
, J e.,.......... -' ~ ~ ' ' ~ ^ * -~ 9?87 PVNGS FSAR '//> ~ FIRE HAZARDS ANALYSIS 4. A deviation is requested fror.. Section III.G.2 to the extent that it requires installation of a one-hour fire rated barrier in addition to an area-wide suppression system. 13 Discussion: The enst wall of the MSSS is non-rated above elevation 100'0" There is a second non-rated wall which abuts the Turbine Building from approximately elevation 110'0" to elevation 140'0a l15 and establishes two void spaces adjacent to Zones 74A and 74B, respectively. Each void is approximately 10 feet by 20 feet and has no combus.tibles. No detection or suppression ic provided within this void space, but both Zones 74A and 74B have automatic suppression and detection. (For additional information refer to the Appendix 9A response to Question 9A.121). 12
== Conclusion:== The existing design provides equivalent protec-tion to that required by Section III.G.2, and upgrading the existing design to a 1-hour rating would not significantly enhance the protection /#J&27-currently provided. bd'% X.
- 5l&
7*bla See the section 9B.2 introduction for generic go deviations. l13 April 1986 98.2.1.2-11 Aniendment IS
cM U T YG 5. A deviation is requested from Section III.G.2 to the extent that it requires fire barriers used for the purpose of redundant circuit separation to be of a " Fire-Tested" configuration. DISCUSSION: Fire Zone 73 contains train A associated raceways protected with TSI thermo-lag fire barrier insulation. The fire barrier insulation was installed as a 3-hour fire-rated system in accordance with the vendor's specification in effect at the time of installation. The fire test performed to qualify the 3-hour rating did not address intervening steel. Further research by the vendor has identified the need for intervening steel which penetrates the protective envelope to be wrapped for 18 inches from point of contact with Thermo-Lag insulation. Since the 18-inch criteria did not exist at the time of installation, the additional protection was not provided. An 8-inch criteria, hoverer, was utilized for installation purposes for intervening steel which came in physical contact with the protected raceway. The 8-inch's of additional protection was measured from point of contact with the protected raceway. The 8-inch criteria, however, was not applicable to intervening steel which penetrated the protective envelope but did not come in physical contact with the protected raceway and associated supports. A " worst case" heat transfer calculation has been performed to determine the minimum fire resistance provided to raceways by the existing thermo-lag configuration. It has been found that a minimum of 18 minutes of protectica will be provideci by the " worst-case" existing configuration. The " worst-case" configuration was found to be P-1000 channel unistrut fully exposed by the fire environment with no consideration for heat sinks. Inis configuration is rare in relation to typical plant installations. Typically the supports are attached to imbed plates or large steel beams which will significantly absorb heat. For example, the duration of protection is increased 50% to 27 minutes if the unistrut support is attached to an imbed plate. The following conservative factors were utilized in the calculation. 1. The configuration evaluated is considered the " worst case" with respect to actual plant conditions. Intervening steel of larger i mass and attachments to large heat sinks such as imbeded plates and I steel beams will significantly increase the duration to achieve 325*F. l l 2. Ine assumption is made that the intervening steel is sub jected to l direct flame impingement with a maximum view factor. Protected l raceways are typically locatert along walls or no high ceilings and direct flame impingement is not linely. In addition, the magnitude l' of radiant and convective heat energy received by the intervening. steel is inverrely proportional to the distance from tte source-9066C/2192A IL-
n , ~ ~.. s-ppgf 5/g [3. The assumption is. made that no heat is absorbed by the thermo-lag through the mechanism of sublimation. Thermo-lag insulation will absorb approximately 750 BTU's per pound in the solid phase and as much. as 6000 BTU's per pound in the vapor phase by endothermic -decomposition.
- 4. -
The assumption is made that all heat received by the intervening V steel is absorbed by the-protected ' raceway conduit with no consideration for re-radiation of heat to the environment. 5. The assumption is made that cable temperature within the protected raceway conduit is equal to the intervening steel. support L temperature.at the point of contact. I E1cetrical cables used for safe-shutdown circuits are IEEE 383 qualified L and meet an additional criteria of resisting 210,000 BTU /llr of neat for the flame test. In addition, our review of PVNGS accelerated aging tests show the ability, to maintain cable integrity even if subjected to temperature extremes at or below 400*F over the 40 year life of the ~ plant. These characteristics, therefore, will provide additional conservatism to the heat transfer calculation discussed above. Fire Zone 73 contains minimal fixed combustibles and has an equivalent. fire severity of two minutes. In addition, the calculated fire loading includes anticipated transient combustibles. The existing Thermo-Lag configuration, therefore, will provide more than adequate protection considering the combustible loading. This Fire Zone is provided with a smoke detection system, Fire liose Station, and portable extinguishers. The Plant Fire Department response is expected within 15 minutes of the alarm condition. l l CONCLUSION: The existing design provides equivalcat protection to that required by Section 111.G.2, and modifying the existing thermo-Jag fire barrier system to comply with the current installation criteria wveld not significantly enhance tr.e protection currently provided. != .& k
3 g 9227 6. A dev!ation is requested from Section III.G.2 to the extent that it requires-fire barriers used for the purpose of redundant circuit separation to be of a " Fire-Tested" configuration. DISCUSSION: Fire Zones 74A, and 74B contain safe shutdown raceways protected with TSI l Thermo-Lag fire barrier insulation. The fire barrier insulation was l installed as a 1-hour fire-rated system in accordance with the vendor's specification in effect at the time of installation. The fire test performed to qualify the 1-hour rating did not address intervening steel. Based. on a three hour fire test, the vendor has identified the need-for intervening steel which penetrates the protective envelope to be wrapped for 18-inches from point of contact with Thermo-Lag insulation for: both 1-hour and 3-hour application. Since the 18-inch criteria did not exist at.the. time of installation, the additional protection was not provided. An 8-inch ' criteria, however, was utilized for installation purposes for intervening steel which came in physical contact with the protected raceway. The 8-inch's of additional protection was measured from point of contact with the protected raceway. The 8-inch criteria, however, was not applicable to intervening steel which penetrated. the protective envelope but did not come in physical contact with the protected raceway and associated supports. A " worst case" heat transfer calculation has been performed to determine the. minimum fire resistance provided to raceways by the existing thermo-lag configuration. It has been found that a minimum of 18 minutes of protection will be provided by the " worst-case" existing configuration. The " worst-case" configuration was found to be F-1000 channel unistrut fully exposed by the fire environment with no consideration for heat sinks. This configuration is rare in relation to typical plant installations. Typically the supports are attached to imbed plates or large steel beams which will significantly absorb heat. For example, the duration of protection is increased 50% to 27 minutes if the unistrut support is attached to an imbed plate. The following -conservative factors were utilized in the calculation. 1. The configuration evaluated is considered the " worst case" with respect to actual plant conditions. Intervening steel of larger mass and attachments to large heat sinks suen as imbed &ed plates and steel beams will significantly increase the duration to achieve 325'F. 2. The. assamption is made that tne intervening steel is subjected to direct flame impingement with a maximum view factor. Protected raceways are typically located along walls or on high ceilings and direct flame impingement is not likely. In addition, the magnitude of radiant and convective heat energy received by the intervening steel is invercely proportional to the dist.ance from the source, f 9066C/2192A
{ :.- s- - *' w D g7 7f 6 3. The assumption is made that no heat. is absorbed by the - Thermo-Lag through the mechanism of ' sublimation. Thermo-lag insulation will absorb approximately 750 BTU's per pound in the solid phase and as much as 6000 - BTU's per pound in the vapor phase by endothermic decomposition. 4. The assumption is made that all heat received by the intervening steel is absorbed by the protected raceway conduit with no consideration for re-radiation of heat to the environment. 5. The assumption is made that cable temperature within the protected raceway conduit is equal to the intervening steel support -temperature at the point of contact. Electrical cables used for safe-shutdown circuits are IEEE 383 qualified - and meet 'an additional criteria of resisting 210,000 BTU /Hr of heat for the flame test. In-addition, our review of FVNGS accelerated aging tests show the ability to maintain cable integrity even if subjected to temperature extremes at or below 400*F over the 40-year life of the plant. These characteristics, therefore, will provide additional conservatism to the heat transfer calculation discussed above. The MSSS Building combustible loading ' consists of cable insulation, Fyrquel oil for main steam isolation and feedwater isolation valves, and transient combustibles. In comparison the oil contained in the valves accounts for the greatest combustible load. All four isolation valves including associated oil reservoirs and tubing are seismic category I qualified components. Based on the seismic design it is reasonable to assume that total leakage of oil from all valves will not occur simultaneous with a fire event. This position is consistent with precedenr.e set for seismic qualified reactor coolant pumps. Given the assumption that the postulated fire will only involve random leakage from the largest single valve, the equivalent fire severity for each zone is 16 minutes. In addition to fixed combustibles, this also assumes 50 lbs. ~ of ordinary combustibles and 55 gallons of oil as transients. In addition "Fyrequel", which has a 650*F firepoint, will significantly reduce the potential for fire ignition and flame propagation should a random leak occur. Due to the low fire loading and the conservative characteristics of the heat transfer calculation for the existing thermo-lag fire resistance, adequate protection will be provided to assure one safe shutdown path will be available. 4 The MSSS Building has a unique roof design which will eliminate the buildup of a hot gas layer. The building nas an open roof design with an 8.5 foot high opening between the top of the wall (elevation 156'-0") and the. botton of the nJssile snield which allows tnc structure to vent prensures developed Juring postulated bign energy line brecks. In addition, the east vall contair.s lari;e openings for main steam and feedwater piping penetrations unicn will allow additienal venting of not gares and products of combust. ion. The physical characteristics or t rie bnilding, thercfore, will prevent room temperatures irom reacning those specified in the ASIM C-119 time temperature curve. i
+ PcP87 Ws-i Fire Zones 74A and 74B are fully protected by automatic pre-action sprinkler systems. Area wide protection, is provided for the 100', 120', 120' mezzanine, and 140' elevations. Each of the elevations above 100' is separated by open grating which will allow some cumulative effect of sprinkler system discharge since each zone is hydraulically designed to allow two elevations to operate simultaneously. Fire detection is provided on each elevation for early warning and actuation of the pre-action valve. Due to the high density of sprinkler heads (65 sq. ft. coverage per head per elevation) it is reasonable to assume that intervening steel will receive direct water impingement and thermal shorts to the protected raceway will be prevented. Both fire zones are provided with hose stations and portable extinguishers for secondary fire supp ession. The plant fire department response is expected within 15 minutes of the alarm condition. CONCLUSION: The existing design provides equivalent protection to that required by Section III.G.2, and modifying the existing Thermo-Lag fire barrier system to comply with the current installation criteria would not significantly enhance the protection currently provided. 1 1 1
JD81 PVNGS FSAR 9g 10bERT tanA.w6) q FIRE HAZARDS ANALYSIS g 10 See section 9B.2.16 for.3 deviation common to Fire Area XVI, section 9B.2.17 for a deviation common to Fire Area XVII, and the section 9B.2 introduction for generic deviations. t C August 1984 9B.2.15-21B Amendment 13 {13
.s ', 2
- 987 MQ 9.
A deviation is requested from Section III.G.2 to the extent that it requires fire barriers used for the purpose of redundant circuit separation to be of a " Fire-Tested" configuration. DISCUSSION: Fire Zones 37B, 37D on elevation 70'-0", and 39B on elevation 88'-0", contain train A associated raceways protected with TS1 Thermo-Lag fire barrier insulation. The exposed train B cables 'and equipment are associated with essential cooling water and safety injection systems. The train A protected circuits are associated with auxiliary feedwater and steam generator cystems and the redundant train B cables and equipment are not located on these elevations. The fire barrier insulation was installed as a 3-hour fire-rated system in accordance with the vendor's specification in effect at the time of installation. The fire test performed to qualify the 3-hour rating did not address intervening steel. Further research by the vendor has identified the need for intervening steel which penetrates the protective envelope to be wrapped for 18-incl (s from point of contact with Thermo-Lag insulation. Since the 18-inch criteria did not exist at the time of installation, the additional protection was not provided. An 8-inch criteria, however, was utilized for installation purposes for intervening steel which came in physical contact with the protected raceway. The 8-inch's of additional protection was measured from point of contact with the protected. raceway. The 8-inch criteria, however, was not applicable to intervening steel which penetrated the protective, envelope but did not come in physical contact with the protected raceway and associated supports. A " worst case" heat transfer calculation has been performed to determine the minimum fire resistance. provided to raceways by the existing thermo-lag configuration. It has been found that a minimum of 18 minutes l of protection will be provided by the " worst-case" existing configuration. The " worst-case" configuration was found to be P-1000 channel unistrut fully exposed. by the fire environment with no consideration for heat sinks. This configuration is rare in relation to typical plant installations. Typically the supports are attached to imbed plates or large steel beams which will significantly absorb heat. For example, the duration of protectiv2 is increased 50% to 27 minutes if the unistrut support is attached to an imbed plate. The following conservative factors were utilized in the calculation. 1. The configuration evaluated is considered the " worst case" with respect to actual plant conditions. Intervening steel of larger mass and attachments to large heat sinks such t s imbeded pir.tes and steel beams will significantly increase the duration to achieve 32T F. 2. The assumption is made that the intervening steel is subjected to direct flane impingement with a naximum view factor. Protected raceways are typically located along we.11s or on hign ceilings and direct flaca impingement is not likely. In addition, the magnitude of radfant and convective heat energy received by the intervening steel is inversely proportional to the distance from the r,ource. 9066C/2192A
s-
- 3 gcf.
"8 // 3. The assumption is made that no heat is absorbed by the thermo-lag through the mechanism of sublimation. Thermo-lag insulation will absorb approximate 1y' 750 BTU's per pound in the solid phase and as much as 6000 BTU's per pound in the vapor phase by ~ endothermic decomposition. 4. The assumption is made that all heat received by the intervening-steel is absorbed by the protected raceway conduit-with no consideration for re-radiation of heat to the environment. 5. The assumption is made that cable temperature within the protected raceway conduit is equal to the intervening steel support temperature at the point of contact. Electrical cables used for safe-shutdown circuits are IEEE 383 qualified and meet an additional criteria of resisting 210,000 BTU /Hr of heat for the flame test. In addition, our review of PVNGS accelerated aging-testa show the ability to maintain cable integrity even if subjected to temperature extremes at or below 400*F over the 40-year life of the plant. These characteristics, therefore, will provide additional conservatism to the heat transfer calculation discussed above. l Fire Zones 39B, 37D, and 37B contain minimal fixed combustibles and have an equivalent fire severity of
- one, four, and five
- minutes, respectively.
In
- addition, the calculated fire loadings include anticipated transient combustibles.
Ine existing thermo-lag I configuration, therefore, will provide more than adequate protection considering the combustible loading. All three Fire Zones'are provided with smoke detection systems, Fire Hose Stations, and portable extinguishers. The Plant Fire Department response is expected within 15 minutes of the alarm condition. CONCLUSION: The existing design provides equivalent protection to that required by Section III.G,2, and modifying the existing thermo-lag fire barrier l system to comply with the currer.t installt. tion c u teria would not I significantly enhance the protection cur'rently provided, L
r safety analysis, and this change further explains 'the results. The operation or design of plant equipment does not change, and the change does not alter the assumptions used in the safety analyses. (72) Description This change ' revised the maintenance program for the switchyard, described in FSAR Section 8A.5, to reflect current practice. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The maintenance program was updated to be less prescriptive, but allowing flexibility to meet the needs of the switchyard. Maintenance testing and inspection will still be performed. The requirements of 10CFR50, Appendix A, CDC 18 and the surveillance requirements of the technical specifications will continue to be met. f (73) Description FSAR Section 9A.68 was revised to delineate requirements for fire barrier penetration seals that cannot be installed in the same manner as tested. The requirements are consistent with the NRC guidance found in Generic Letter 86-10. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. The penetration seals will be evaluated to ensure that an equivalent level of protection will be provided, and the barriers will be maintained to prevent the spread of fire. The installations must meet the NRC guidance found in Generic Letter 86-10. In addition, backup measures such as detection and suppression are provided in case of seal failure. (74) Description FSAR Section 9.1.3.3.1.1 was revised to reflect the lowering of the spent fuel pool high level alarm PCN-LSHL-003 from elev. 138'-6" to 138'-2". This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safety Evaluation This change did not intreduce an unreviewed safety question. Reducing the distante between the normal pool level ans eh* high alarm restricts the fluctuation of tha water levs1 to a r.maller band, producing earlier warning of potential problems. The function of the alarm has not cht.nged. .#.J// gh FILE: T/- C3 3 'lDN PAGE ! 0F 10CFR50.59 REVIEW AND EVALUATION ACTION UNDER REVIEW: L REVISION: DESCRIPTION OF PROPOSED CHANCE:b mo Y/ f> b[ h / M /F R OOu[M/k?My cY))N9 WDRb,9 j Ay) Tag b Y? () (JW)/) (/)k o h nhpt L% sad mum run Md. 10CFR50.59 REVIEW DOES THE PROPOSED CHANGE: 1. MAKE CHANGES IN THE FACILITY? 2. MAKE CHANGES IN PROCEDURES AS THEY ARE DESCRIBED IN THE FSAR?.YESd NO YES / Nog 3. INVOLVE TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR? YES NO / 4 INVOLVE ANY OTHER CHALLENGES TO NUCLEAR SAFETY FOR PVNGS? YES NO / 5. REQUIRE A CHANGE TO THE TECHNICAL SPECIFICATIONS? YES NO,, / 10CFR50.59 EVALUATION (Provide Response Justification with References) 6. WILL THE PROBABILITY OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES NO / THE FSAR BE INCREASED? 7 WILL THE CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED IN YES NO [ THE FSAR BE INCREASED? 8. WILL THE PROBABILITY OF A MALFUNCTION OF EQUIPMENT IMPORTANT YES NO [ TO SAFETY BE INCREASED? 9. WILL THE CONSEQUENCES OF A MALFUNCTION OF EQUIPMENT IMPORTANTYES NO if TO SAFETY BE INCREASED?"
- 10. WILL THE POSSIBILITY OF AN ACCIDENT OF A DIFFERENT TYPE THAN YES NO /'
ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 11. WILL THE POSSIBILITY OF A MALFUNCTIONING OF A DIFFERENT TYPE YES NO [
THAN ANY PREVIOUSLY EVALUATED IN THE FSAR BE CREATED?
- 12. WILL THE MARGIN OF SAFETY AS DEFINED IN THE BASIS FOR ANY YES NO [
[TECHNICALSPECIFICATIONBEREDUCED? ~ ANY ANSWER TO QUESTIONS 1 THROUGH 4 "YES", THEN A 10CFR50.59 EVALUATION IS REQUIRED. FSAR CHANGE REQUEST PER PROCEDURE 5N404.01.00 MAY ALSO BE REQUIRED. ANSWER 5 IS "YES", THEN TECHNICAL SPECIFICATION CHANGE REQUEST PER PROCEDURE SN404.01.00 AND NRC APPPOVAL REQUIED PRIOR TO IMPLEMENTATI ANY ANSWER 10 QUESTIONS 6 THROUGH 12 "YES" THEM AN UNREVIEWED SAFETY ~~ QUESTION IS IDENTIFIED FROCEED TO PPOCFDUiE 7N407.02.00 PRIOR TO IMPLEMENTATION. / Ali ANSI,ERS 6 THRU 12 ARE "NO" REC 03NEND ACTION APPROVAL.
- LL ANS'.iERS 1 THROUGH 5 ARE "NO", NO 10CFR50.59 EVALUliTION REQUIRED,
~ RECOMMEND AC MON APPROVAL. I verify t.htt the above review / evaluation is adect: ate M nccurate and that at least ore of the undersigned has received required training. ,l X jm/ 'c m Y y
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s.- [ TVNGS FSAR l Jg APPENDIX 9A ( used in three-hour walls and 1-1/2-hour dampers are used in two-hour and one-hour walls. The fire ' dampers purchased for PVNGS are all of identical 13 material and constructed to 3-hour standards. (Refer to the response to Question 9A.108). Class A doors are used in three-hour fire walls, [F class B doors are used in two-hour fire walls and 13 Class c doors are used one-hour fire walls. (Refer to the response to Question 9A.106). c. Fire barrier penetration seals fr Testing and acceptance criteria are as specified in ASTM tandard E-119 and IEEE 634 (1978). Seals 1 /Co_l 105e.n Ge,lity ar inst led in the same manner as tested.~ Qua 13 assurance, quality control and other measures are made to insure that the actual installation conforms to the specified requirements. The cable trays are / . supported by tray supports located close to the wall penetration to increase the reliability and 7 integrity of the raceway system in case of fire. Consequently the penetration seals will not be affected due to unsupported load. (Refer to the 13 response to Question 9A.110). d. Metal deck roof All roof slabs in safety related areas are of p structural concrete. Structural thickness exceeds the three-hour fire separation requirements. QpESTION 9A.69 (FPER Question 7) (9a.3) Page III-15 *), Item D.1.(i): Verify that the floor drains I 13 are of adequate size to handle any run-off from any water i ( a. Page references are no longer applicable due to FPER reformatting for FSAR Amendment 13. August 1984 9A-55 Amendment 13 h3
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4 c. (78) Description FSAR Section 17.lB was revised to reflect that ANPP would retain and maintain original design documentation, as opposed to it being part of Bechtel's Quality Assurance (QA) program. The documentation will be maintained in accordance with the ANPP QA operations program for document control and retention. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The commitments 6f record retention continue to be met, just by a different organizccion. The documentation continues to be maintained and is j retrievable for reference when making design decisions. (79) Description FSAR. Section 8.3.1.1.4.6 was revised to clarify the description of the load shedding circuits. In addition, Table 8.3-4 was updated to reflect the 120VAC Vital Power System Loads. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. The additions to the 120VAC Vital Power System Loads ensure that the operators have the necessary instrumentation to respond to an accident condition. The change to the load shed description clarifies that the primary and secondary breakers remain closed on a load shed signal to avoid overloading the breakers. The changes are made to enhance equipment performance and/or availability and do not impact the ability of the equipment to perform its intended function. (80) Description FSAR Section 17.2.6, on Quality Assurance document control, was revised to clarify that each department is responsible for the preparation,
- review, approval and distribution of its own departmental directives, guidelines, and instructions. This change was transmitted to the.NRC in the USAR, Rev. O.
Summarv of Safetv Evaluation This char.ge did not introduce an unreviewed safety question. The change was administrative, to clarify departmental responsibilities. { It did not result in a reduction of QA commitments. t (B1) ILeJp_ripti on FSAR Table 7A-3 wu reviseo to reflect that the Aw.iliary Feedwater Accuation Signal (APAE) ?ogic vas changed to remove the "ATAS-1 i l i
r. 1 J I' s 1 priority over AFAS-2" feature of the auxiliary feedwater pump. turbine steam supply valve. In addition, the manual override was deleted. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safety Evaluation This change di_d not introduce an unreviewed safety question. The change will not impact the containment post accident equipment qualification temperature. The purpose was to minimize operator error and improve operational flexibility. (82) Description FSAR Section 9.2.8.5 was revised to reflect that calcium analyzer
- AIT-075 was abandoned in place, and that chlorine detection instruments (calcium appears in fixed proportion to chlorine and is used to measura chlorine) are no longer provided to detect inleakage from the plant cooling water system (PCUS) to the turbine cooling water system (TCWS). This change was transmitted to the NRC in the USAR, Rev. O.
Summarv of Safety Evaluation This change did not introduce an unreviewed safety question. The TCWS is not an important to safety system and is not credited in any FSAR accident analyses. A malfunction of the TCWS would not affect any system or component important to safety. A grab sample point downstream of the TCWS pumps provides for periodic water analysis. (83) Description FSAR Sections 12.5.2 and 12.5.3 were revised to clarify information regarding radiation protection activities and accurately reflect dcsimetry calibration and reading intervals. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Safetv Evaluation This change did not introduce an unreviewed safety question. No changes are made to equipment or systems important to safety and procedures for safe shutdown or accident mitigation are not affected. The calibration checks for the radiation monitors are governed by technical specification requirements. (84) Description FSAR Section 8.3.1.1.6 was revised to delete the requirement to replace the manual transfer switch for the class inverters with static transfer switeben prior to restart from the first refueling outage. This change was submitted to the NRC by letter dated 10/15/87 (161-00589). and also in the USAR, Rev. O. _ _
~.,_*".:.. I').\\C-o%2W/ 3 10CER50.59 Review and Evaluatlon b ACTIDJLatER REVIEUs W a-M E f *A / g eie.7).,9 REVISICh DESUtIPTICJ 0F FRG305ED CHANCE: ' C, - C-0$ns/C F We> ESA S lcA hr C 2$ (Yxn/YW v r st DCP 'S ' 10CTR50.59 REVIEW (PrevidePeferencesonResponseJustificationPage) NO YES Does the proposed change: 1. Make changes in the facility? 2. flake changes in precedures as;they areidescribed in the FSAR7 / 3. Invo2vetesterexperimentsnotdescribedintheFSAR? [_ 4 Recuire a change to the technical specifications? g_ Any answer to cuestions 1 through 3 "YES", then a 10CFR50.59 evaluaticn is re F5AR Change Recuest per procedure SN404.01.C0 may also be recuired. quired. Ansver 4 is "YES", then Technical Specification Change Recuest per precedure 5:204.01.00 and NRC espreval is required prior to implementation. All answers 1 through 4 are "NG", no 10CFR50.59 Evaluation required. rece nend acticn acpreval. 10CFR50.59 EVALUATI2J (previde Response Justification with References)
- 5. Uill the probability of an accident previously evaluated in the FSAR te increaseo?
6. Will the consescuences of an accident previcusly evaluated in the FSAR te increaseo? 7. Uill the pretability of a malfunction of equipment important to safety te increasea? 8, Will the consequences of a malfunction of equipment important to' safety be increasee? 9. Will the possibility of an accident of a different type than any crevicusly evaluated in the FSAR be created? 10. Will the possibility of a malfunctioning of a different type than any previcusly evaluated in the FSAR be createe? 11. Will the margin of safety as defined in the basis for any technical scecificatien be reduced? Any ansuer to questiens 5 thrcuch 11 "YES". then a potential enrevie.ed safety cestien is identified. Prc:eed to cricecure 7: All anm.ers 5 threcgn 11 are "NO". rect reno acticn a steval.KD?.C3.C0 prict to i :lt ents:!c _ If FSAR chacter S/Chacter 15 is actenti.111y of factcc fc:uste a c::.y cf avelvatica to ??uclear Feels ranagement. I verify that Ebg ahove revic;/evaluatica 13.tdequate ard oc;urstz ord thrt the LT10e ( re civce.1 required training, Aes s $d 9bd? /E/~ 2 ome Vh3b7
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"A N8O y 107R50.59 REVIEW AND EVALUATION I RESPONSE JJSTIFlCAT10N PAGE3 CF / W ACTION UNDER REYlEW: MM8CA/ gefMJ PIV ISION: 8 Name/Ti t le FROCEDURE/PCP/TEF'P MOD. NO: A/M I CUESTICN I RESPONSE JUSTIFICATION LTY fdeS 7 Y88 7 ).-}f f - Q [ ] pp' f F F F A /e* f s* ** I I 1 I I I i 1 1 1 1 I i i i I i i l i I I i l I i i l I i i l i I__ 1
fh gh 10CFR50.59 - SAFETY REVIEW / SAFETY EVALUATION 8 N O M 2 Rev. [ No. Initiating Docu:nent ~2 7 W Description of Proposed Change: IH 52c A 7. 4/%hx 7A Tah/.e 7A - 3, & & 3 :i f C2!4S D 0 T0 v l vt. ty' y l*'t t IT h l/& t EACH QUESTION BELOW SHALL BE ANSWERED ON AT 10CFR50.59 Review toes the proposed change: YES NO 1. Make changes in the facility? 2. _V Make changes in procedures as they are described in the FSARf g 3. Involve tests or arperiments not described in the FSARf 4. Involve any other challenges to nuclear safety for PYNCS? / 5. Require a change to the technical specifications Y ._ v' i 10CFR50.59 Evaluation ) 6. Will the probability of an acci5ent previously evaluated t/ in the FSAR be increased? ~ 7. Will the consequences of an accident previously evaluated / in the FSAR be increased? 8. Will the probability of a malfunction of equipment 7 important to safety be increased? 9. Will the consequences of a malfunction of equipment / important to safety be increased? 10. Will the possibility of an accident of a different type than any previously evaluated in the TSAR be created? y 11. Will the possibility of a malfunctioning of a different type than any previously evaluated in the FSAR be created? _t/ 12. Will the margin of safety as defined in the basis for any / technical specification be reduced? b If any answer to 1 through 4 is "Yes", an FSAR Change Notice (SARCN) may be required.* (SARCN #_ M M )** NO FSAR Chaptar 6/ Chapter 15 analyses potentially af fected? Overall system performance shown to meet requiramants of safety analysist== NO If answer to 5 is "Yes". Technical Specification change request and NRC approval required prior to 3M lamentation.** N0 If any answer to 6 througn 12 is "Yes", an Unreviewed Safety Question, exists. Reviewer:,4 6 & #- M FRC kpproval required prior to implementation.** Data 3/f7 9!~*o NCS approval: 'ths Date ') y e /
- Tamporary modifications, temporary procedures, or special (one time only) tests do not require an FSAR che.nge.
- If response is "Yes," the '3echtel NCS will contact the ANPP Puclose Licensing Manager.
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a. ~ 10CT1tS0.59 EAFETY tryIEg AyD IVALUAUDN C RESPoESE." JUSTIFICATION N -8 M d,,,, [ 50. Page ~7 og { Initiating Document INN Eqq7 I i l I _wrsnes 1 trsrevsr JusTtFICAnot 1 l { ll l Th-e. fasscinli/y a f a. A.L a ac Ic i &l 1 i cllLLA.t %f* EL)9 G lkA, & Stsvtbatlv l l t L ~ l l.e v s L 'a k) k v4s Fsn-a A ll N I-b.e l l lev.uth) 5:a u TLa Dc.Ps o< voletd 4 l l l /o 9ic e 4.t.Mcr aJ uo odJ<k M l l l waMr aa eM:.).. Thir yropas-e) l l l e-- Q t tail / %ke FM/2. 7. &/-L 7 4.3 l l j h ic/Le 4 w < M U.o Accth d 9 .s_r/g. l l /2 ! Th.it #<0por-e) 24 aa-c Aas u bpacf-l l l fo is spiilte able LR-c.fuk of 4 fe-d. Ve h c l lf% e.f-or.s O m w e'n of Salely asdakal l \\ in. %p b$ sis (~w any bN-c4. Lac.s. w i ll ' }. % Y' bb Y LS% C.C i I l l t l I I 1 .. _ = - i 1 I l i J l 1 I I I _I I I II ~ I t 9 o &~) g, 2585r 33_ g
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o Summary of Safetv Evaluation The change did not introduce an unreviewed safety question. The change did not alter the design of Unit 1 or assumptions made in the FSAR safety analyses. The current manual transfer switches comply with all regulatory requirements. o (85) Descrivefon FSAR Sections 11.5.2, 12.5.2 and Table 9.3-3 vere revised to clarify the description of radiation monitoring equipment, location, operation, and calibration criteria. This change was transmitted to the NRC in the USAR, Rev. O. Summarv of Saferv Evaluation This change did not introduce an unreviewed safety question. The changes can improve the ability of Radiation Protection personnel to respond to and mitigate the consequences of an accident. 4 (86) Description FSAR Section 7.4.1.1 was revised to reflect changes to the undervoltage protection system for the 4.16kV ESF bus. In response to an NRC requirement in the PVNCS SER, an additional protective trip was inst 411ed at approximately 90% of nominal bus voltage, with a time delay to avoid spurious trips due to short duration transients. This was in addition to the 70% voltage setpoint used to detect loss of offsite power. Also, FSAR Section 7.4.1.1 was clarified to accurately describe the permissive signal to the diesel generator automatic starting sequence. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Saferv Evaluation a This change did not introduce an unreviewed safety question. Because offsite power will be tripped and the diesel generator started on a higher undervoltage, potential damage to equipment due to a sustained undervoltage vill be prevented. The diesel generator will not be impeded from starting daring a loss of offsite power, nor have the trip setpoints changed. This changs was found acceptable by the NRC in PVNCS SER Supplement 5. (87) Dnerfyhn FSAR Section 9B was revised to reflect the actual thermolag installation in fire zon2s 10B. 423, 41C, 46A, 46B, 46E, 47B. and 52D. This represents a deviation from Section III.C.2 of Appendix R to 10CFR50. This chan6e was transmitted to the NRC fn the USAR R9v. O. E.g aru of Safetv Evaluation This change did not introduce an unreviewed safety question. The existing installation is equivalent to that required by iec: ion III.C.2 of Appendix R. _______-_-_-___-___ -___ - -_ -
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~ c, 10CFR50.59 Review and Evaluation %\\Q l ACTION L;; DER REVIBJs REVISION: l QA dc4 cmc 0 "Tb toep.40, @ 5eccan ni g.t, 7%R g3nga q6,"L gh ~ DESCRIPTION OF PROPOSED CHANCE: W Ed ann k km ~2Ones 10%416,07f bh bh 4(eE,4]6 Wd 6ED prt Fute "E x o cn_ M-catT Msm\\Wtwoa. 10CFR50.59 REVIDI (Provide References on Response Justification Page) PO YES 1 l Does the proposed change 1. Make enanges in the facility? / 2. Make changes in procecures as they are cescribed in the FSAR? / l 3. Involve test er experiments not described in the FSAR1 / 4. Recuire a change to the technical specifications? / / Any ensuer to cuestions 1 through,! "YES", then a 10CFR50.59 evaluaticn is recuired. F5AR Change Recuest per precedure SN404.01.C0 may also be recuired. Ansuer 4 is "YES", then Technical Specification Change Recuest per procecure SN404.C1.00 and NRC approval is recuired Srior to implementation. All ansvers 1 thrcugn 4 are "!D", no 10CFR50.59 Evaluation recuired, rec: mend acti:n a;::::: val. l 10CFR50.59 EVALUATION (Previde Response Justification uith References) l 5. Uill the pre:acility of an accident previcusly evaluated in the FSAR te increaseo? / l ~ E. Will the c nzescuences of an accident previcusly evaluated in the FSAR be increasec? / 7. Will the probability of a malfunction of eculpment important to safety te increasee? / 8. Will '.he censecuences of a malfunction of equipment important to safety be increaseo? / 9. Will ne possibility of an accident of a dif ferent type than any previously evaluated in tre FSAR te createc? / l
- 10. Will the possibility of a malfunctioning of a cifferent type t".an any previously evaluated in the FSAR te created?
/
- 11. Will the margin of safety as defined in the tesis for any tecnnical t:ecifi:atien be reduced?
/ 1 Any ansuer to cuestiens 5 through 11 "YE5", then a potential unrevieued safety cuestien is identified. Prceeed to prececure 7m.07.03.00 pri:r to irplerentati:n. All ansvers 5 througn 11 are "!0", re : mend a:ti n 3:::reval. l If F5AR chapter 5/ Chapter 15 is potentially of fectec, forwarc a cc:y cf evaluatien to fluclear Fuels Manage ent. I verify that the above revicv/ evaluation is acc<pate and accurate and that the unocreigned baue received recpired training. 9-lD-87 00 " _9 r yms -
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h-l: cPJ63 ~ 10CFR50.59 REVIEW AND EVALUATION RE!PONSE JUSTIFICATION 0F N PAGE 1 ACTIGN UNCER REYlEW: $kQ Cwoom s) blL b+cwds hodsd M REYl 510N: hh tieme/ Title L PRCCEDURE/PCP/TEFP N00. NO: OfA I OUESTION l RESPONSE JUSTIFICATION I k. IThs CAMOhE M4fis A cmEE tu THE FAcitirv M M<crise b !J l i Fsaa. Lucclir es -m veheex -me-Ac-ro at '1h e o s o - L a 3, m tis n m I Iio be- +enes toe [ A27, A 2_c 4toA,406 Abf Att, ed 62.D. I 7, l%s ckwe .s Oct MESC/ibed to hoM wetedu/es swr 1 IT n ro reMeer t i l me As-Eoser 10srntta-ned new -Tweneec me meneus os dcccoun dleered. i tu rm F5NL w e. ocv I 3 i%s FSML C A c-,E. DOES GOT lOVoiVE InTS 00 Euc<ugears ucrr I l'bg.qn_t tt, c.t> 10 -TWE Mi '6 eta 0'.:f-M 15 A bfVtATbd m t odff 50 / ILoot, Set. sea E.L>.~L oda.* ~TMM Ne uo -rtsis ok. 6xwunws a s i Ve rovdue. mke, I6c6Aa.\\ 4. I 6ft &crrEunog b/(S We Ocn-Dest /t 40 l I bx-t Lb-Do05. I S. th FSAL Acradeer Eudomeo AOccad 'N mn cacoc.c udAd Lc. h/c, worwnoa eVdornod ' k W 'r APOMm s chawo c. d ThE I I OcyT ldC# CASE. M (obA M A /C-QCO / nug h3tl M s i I GT\\)(c da f.CM f-A DS bV b C. 14 A $#/t def M *T' I o i i 5% C - Sb OW O O AN n r*. tPAC iW / 1 'D/fdGM-r5 P Ye i 55 LVC-. T CW\\ d ) lLi T*Wehyt. 'de h/c. kWWb lebis AM fvt mmrid e k itTiu q 6 i J t Term 7t4407.01B
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'rei ; e=ch fire ::ca durcription, a litailed itemization of
- .re crea spec.'.fic deviations from the separation requirements cf Ecc:i n II.G is also presanted.
In addition, ceveral genert.c deriations with relevance te many of the fire areas are: A. In general, exterior walls, basemats and roofs are not rtted when: 1. They are not required to separate a safe shut-down related train inside the fire crea from a significant fire hazard (e.g., oil filled transformers) outside the fire area, and 13 2. They do not separate safety related areas from ~ non-safety related areas that present a signi-1 ficant fire threat to the safety related areas. The existing design which includes non-rated exterior walls, basemats and roofs is an acceptable alternative to the separation requirements of Section III.G.2 of Appendix R. B. Some fire doors have been modified to include security system hardware. In general, the modifications have a minimal impact on the door fire resistance as the modification affects a limited area of one side of the door. Further, these modifications are necessary to provide adequate station security. The modifications are in accordance with industry practice and are considered acceptable in that without modification the doors would cause a condition detrimental to overall facility safety. C. August 1984 Amendment 13 l13 CB*2*1-1 ~
'L, 'p A351 C. A deviation is requested from Section III.G.2 to the extent that it requires fire barriers used for the purpose of redundant circuit-separation to provide a cold side temperature of less than 325"F on the safe shutdown cable. DISCUSSION Fire Zones 10B, 42B, 420, 46A, 46B, 46E. 47B and 52D contain Train A associated safe-shutdown raceways protected with the - TSI Thermo-Lag fire barrier system. The fire barrier was installed as a 1-hour fire-rated system in accordance with the vendor's specification in effect at the time of installation. -The fire test performed to qualify. - the 1-hour rating did not address intervening steel. Further research by the vendor has - identified the need for intervening steel, which penetrates the protective envelope, to be wrapped for 18 inches : from the point of contact with the' Thermo-Lag envelope. Since the 18-inch criteria did not exist at the' time of installation, the additional protection was not provided. An 8 inch criteria was utilized for installation purposes on intervening steel which came in physical contact with the protected-raceway. The 8-inch criteria, however, was not applicable to intervening steel which penetrated the protective envelope but did not come in physical contact with the protected raceway or associated supports. A fire test has been performed to determine the effect of having less than 18-inches of protection on an intervening steel member in contact. with the protected raceway. The fire test results were then extrapolated in order to determine the effect of 8-inches protection on the intervening steel member. The resultant temperature at the protected raceway was found to be less than-175'F. Therefore, the 8" of protection as presently installed on intervening steel that penetrates the protective envelope and is in physical contact with the protected raceway, is acceptable for a one-hour rating. A " worst case" fire test was also conducted to determine the effect of intervening steel that does not contact the protected raceway and is not wrapped with Thermo-Lag. The test demonstrated that a 400'F temperature could be maintained on the inside of the protected raceway. A review of accelerated aging tests shows the ability of the cable to maintain it's integrity, even if subjected to temperature extremes at or below 400*F over the 40 year life of the plant. The condition tested was a 3/4", schedule 40, pipe two inches away from a two-inch conduit. This l configuration is conser rative in relation to actual plant installations, since the penetrating elements are typically small conduits and ground wires which are attached to imbed plates, large steel beams or concrete walls which would tend to absorb significant amounts of heat. l Additionally, the test fixture was a small box (0.5 cubic feet) which restricted heat dissipation. Field installations of the fire barrier typically enclose a large volume and cover the entire length of the raceway in a particular room. (.- 9622A/2234A __--_ - _ _ _ _ - _ - _ _ a
. S', / CMS lL k Electrical cables used for safe-s'hutdown circuits are IEEE-383 gualified and are capable of resisting a heat flux of 1.76 BTU /sec-ft before sustaining damage. A calculation has been performed that shows that if the clearance between the safe-shutdown cable and the penetrating item is a minimum of 0.10 of the radius of the cable, the penetrating item would have to have a temperature over 1000*F to cause a heat flux sufficient to damage the cable. The above referenced fire test also measured the surface temperature of the 3/4-inch pipe, inside the envelope. Since this test showed a maximum surface temperature of 735'F, a temperature of 1000*F inside the protected envelope is not believed to be credible. Additionally, the calculation is conservative in respect to actual plant installations because it allows the penetrating item to be in very close proximity to the protected raceway (1/10 of an inch for a 2-inch conduit). The above fire zones contain fixed combustibles, primarily in the form of electrical cable insulation, with the exception of zones 46A, 46B and 46E which are the charging pump rooms. These zones are provided with ionization and line-thermal detectors with automatic pre-action systems in the cable trays (Zone 10B has an automatic halon system and Zones 46A, 46B and 46E have area pre-action sprinklers and ionization detectors). The fire zones are also provided with hose reels and portable extinguishers as secondary fire suppression systems and plant fire department response is expected within 10 minutes of the alarm condition. CONCLUSION The existing installation provides equivalent protection to that required by Section III.G.2, and modifying the existing Thermo-Lag fire barrier to comply with the current suggested installation criteria would not significantly enahnce the protection currently provided. 1 __-_-_____-___-_-______-___-_--a
(88) Description FSAR Section 9.5.1 and Appendices 9A and 9B were revised to reflect the as-built condition of the plant. Combustible loading figures were updated to reflect combustibles added as a result of plant changes and occupancy. In addition, reporting responsibilities and department titles were revised to reflect the existing organization. This change was transmitted to the NRC in USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The change does not decrease the fire protection commitments or reduce the leve) of protection as presently described. The change also does not affect separation criteria or spurious actuation analyses and the existing fire hazards analysis remains valid. (89) Description FSAR Tale 11.2-1 was revised to allow the use of 316 stainless steel (SS) in addition to 316L SS for the liquid radwaste (LR) recycle monitor pu=p LRN-P03. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The 316 SS physical properties are compatible with the LR chemistry and are comparable to 316L SS. The pump is a non-safety related, non-ASME Section 3 pump and its electrical or mechanical function is not affected by the use of a higher carbon content in the SS material. (90) Description FSAR Section 8.3 was revised to reflect changes to the Degraded Electrical Power procedure that allow for connecting both ESF buses (503 and SO4) to a single operating diesel generator after a loss of power when one diesel generator is inoperable. This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safetv Evaluation This change did not introduce an unreviewed safety question. The change did not result in any system modification. The effectiveness of operator recovery actions to mitigate a possible core damage event are increased by reducing the amount of time necessary to perform them. In
- addition, the conditional core damage probability is reduced by j
approximately a factor of 2.1 when the revised procedure is utilized. \\ (91) Description l ( FSAR Sections 6.2.4 and 6.2.6 were revised to modify the description of I the containment isolation system to reflect as-built conditions. This I change was transmitted to the NRC in the USAR. Rev. O. ( j l I
\\ a Summarv of Safetv Evaluation This change did not introduce an unreviewed safety question. All changes were in accordance with 10CFR50 Appendix A. Testing per 10CFR50 Appendix J will still occur. (92) Description Question 7A.4 of FSAR Appendix 7A was revised to describe the impact of loss of power to panels E-NNN-Dil and E-NNN-D12 on pressurizer heater operation depending on the position of hand switches HS-100 and HS-100-3.
- Also, the impact on the pressurizer level control and the pressurizer pressure control systems, and the operation of the charging pumps was clarified.
This change was transmitted to the NRC in the USAR, Rev. O. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The loss of backup heaters is within the accident analyses in the FSAR. All actions are detailed in existing plant procedures. (93) Description FSAR Table 3.2-1 and Section 17A.62 were revised to make the following changes: 1. Add ERFDADS to Table 3.2-1 (Quality Classification of Structures, Systems and Components). 2. Add accident monitoring instrumentation to Table 3.2-1. 3. Modify response to questions 17A.62 C.1, 2, 4, 5, 11, 12 and 17 to refleet completed design. 4. Modify footnote u in Table 3.2-1 to reflect Reg. Cuide 1.97 QA requirements. This change was transmitted to the NRC in the USAR, Rev. O. l Summary of Saferv Evaluation 1 This char:ge did not introduce an unreviewed safety question. No changes are made to the facility or procedures. This ch~nge clarifies the QA requirements. (94) Description l FSAR Sections 93.2.12 and 9B.2.15 were revised to make editorial changes and update the combustible loading calculation and fire department response time figures, for consistenc; with the changes reported in item w88 of this report. These changes also clarified the deviations added by item n58. This change was transmitted to the NRC in the USAR, Rev. O. - _ _ _ - _ _ _ _ _ _ _ _ _
.s q 1~ 79AC-9ZZ07/5 )f; l 10CFR50.59 Review and Evaluation
- M ACTIO*J LNDER REVIEW:
TSAR Chance to FSAR 6.2.4 & 6.2.6 REVISION: DESCRIPTION OF PR T OSED CHANGE: To incorporate design changes / equipment changes into FSAR which have occurred previously but have not been reflected in the FSAR. 10CTRSO.59 REVIEW (Provide References on Response Justification Page) NO YES Does the proposed changer 1 1. Make changes in the fscility? X 2. Make changes in procedures as they are described in the FSAR? X 3. Involve test or experiments not described in the FSAR? X l 4. Require a change to the technical specifications? X X Any answer to questiens 1 through 3 "YES", then a 10CFR50.59 evaluation is required. FSAR Change Request per procedure SN404.01.00 may also be required. Answer 4 is "YES", then Technical Specification Change Request per procedure SN404.01.00 and NRC approval is required prior to implementation. All ansuers 1 through 4 are "NO", no 10CFR50.59 Evaluatien required, recommend action approval. 10CFRSO.S9 EVAltlATION (Provide Response Justification uith Referencer) 5. Will the probability of an accident previously evaluated in the FSAR be increased? X S. Will the consequences of an accident previously evaluated in the FSAR be increased? X 7. Will the probability of a malfunction of equipment important to safety be increared? X 8. Will the consequences of a malfunction of equipment inportant to safety be increased? X 9. Will the possibility of an accident of a different type than any previously evaluated in the FSAR be created? X 10. Will the possibility of a malfunctioning of a different type than any previously evaluated in the FSAR be created? X 11. Will the margin of safety as defined in the basis for any technical specification be reduced? X Any answer to questiens 5 thrcugh 11 "YES", then a potential unreviewed safety question is identified. Preceed to prececure 7N407.03.00 prior to implementation. X All answers 5 through 11 are "f.'0", reccmend acticn approval. If FSAR chacter 6/ Chapter 15 is potentially af fected, fcruard a ccpy of evaluation to Nuclear Fuels Managtment. I verify that the above revitv/cvaluation is adequate and accurate and that the undersigned have r - required trai 'ng. fulw suna.wed %6 w w INITIATCR ' CATE cff I.1IT'ATCR'S ELpERV15GR GR S.S. CME
1 T O' e = ab59 p-10CFR50.59 REVIEW AND EVALUATION RESPONSE JUSTIFICATION J. 2 PAGE 0F ACTION UNDER REVIEW: Changes to FSAR 6.2.4 & 6.2.6 REVISION: Name/ Title PROCEDURE /PCP/ TEMP MOD. NO: N/A i _0UESTION l RESPONSE JUSTIFICATION 1 -- g All changes to the facility were covered separately under DCP's or S-Mods. This SARCN only updates the FSAR to reflect as-built condition i l of PVNGS plants. Each design change was separately reviewed per 10CFR50.59. Ti$e changes made to the FSAR by this change are editorial in nature or reflect as-built configuration. All changes to the plant are still in accordance with 10CFR50 Appendix A. Testing per 10CFR50. Appendix J (ie LLRT Type B & C testing 6 ILRT) will still occur per Tech Spec 4.6. Changes to FSAR will ~not change probability or consequences to accident since changes only reflect as-built condition. h I I M 79AC-9ZZ07/5 E _ ___ __ _ _ _ _ _ _ - _. - - _ _ - _ - - -
. 3 0361 ~ N' > 2 PVNGS FSAR CONTAIN YSTEMS or. ydebs. hatch to the annulus between the two 0-rings 4 When assembled preparatory to reactor operation, the blind flanges and hatches are secured to.th ccess connection and the annulus between cr we. s the 0-rings ^'is p<ressurized to ensure that both seals are functioning. The seal is further tested when test pressure is introduced into the containment. s.M,# verst, s r un, g;/j"~.The emergency personnel hatch, personnel lock, equipment ~ /j e - hatch?/ test connection and fuel transfer connection are ~~ considered to be part of the containment boundary and therefore General Design Criterion 56 does not apply to these penetra- 'I tions and an isolation valve is not required. -.y.y y.w 7 .5 c,{hh{.,b/[ 1-[y ' 6.2.4.3 Desian Evaluation x. - -. Refer to CESSAR Section 6.2.4.3 for systems covered under g, CESSAR licensing scope. Single failure analyses of BOP ESFAS Yh9 y - _ a. and NSSS ESFAS are presented as part of section 7.3 and f]{J CESSAR Section 7.2. Single valve failure does not affect W _3 the integrity of the containment building due to redundancy of [rM h* double isolation valve protection. There are two classes of gno exceptions to double isolation valve redundancy. Single isola-Lion is provided for the main steam lines in accordance with the provisions of General Design Criterion 57. Containment pressure instrumentation has single isolation in accordance with Regulatory Guide 1.11. Operator action will be required to isolate a pressure instrument line rupture downstream of the isolation valve. As noted in section 6.2.4.2, the equipment hatch, the emergency personnel hatch, the personnel lock, the containment test 10l cunection, ILRT verificati m, ILRT pressure measurement, and the fuel transfer penetration all have closures surrounding the access pipe with a blind flange fitted 10\\ with double 0-rings or gaskets which serves as the primary containment seal. The respective access pipe closures and f~ y f k ~ .g. ( </_ ,3, , - ~.: .. L_ C l'~-... G/! Amendment 10 6.2.4-24 December 1982
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s I 9359 9f VALVE PENE. ARRANGEMENT INSIDE CONTAINMENT DUTSIDE CONTAINMENT TRATION N O. 30, (17,20) _ PE NETR ATtON M 13,14,15,16 I',.- M 1 SEE NOTE 1 17,18,19,20 =
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44 r, .'. '.N { D. T D, T Palo Verde Nuclear Generating Station FSAR i 1 cc::TA::1: r:;T PE::r:?.AT:c:q VArvt AaaA::ct:ts::: l14 (Sheet 1 of 10) Fignre 6.2.;-l
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s s c93.5 T IOf y A-VALVE PENE-ARRANGEMENT INSIDE CONTAINMENT GUTSIDE CONTAINMENT TRATION l N O. NO.
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1 P A A t'.'. i L ~' y D, T UD W D. T D,T LEGEND: D = ORAIN V = VENT T = TEST IV = ISO LATION VALVE NOTES:
- 1. VALVES ARE OPENED IN EVENT OF LOCA.
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- 2. VALVES ARE IN SECONDARY S10E OF STEAM GENERATOR.
IN THE EVENT 0F LOCA,THE SG TUBES CAN BE WATER SEALED BY FLOODING STEAM GENERATOR, AND THERE. FORETYPE CTESTS ARE NOT PERFORMED.
- 3. DNE BUTTERFLY VALVE WILL BE LEAKAGE TESTED IN REVERSE DIRECTION.
Palo Verde Nuclear Generating Station FSAR Co::TA:::".r::T PE::nTRAT:c:: t/A Vr An?x:crar::: (Sheet 10 of 10) CHANGE
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t 2459 ~ 89f PVNGS FSAR CONTAINMENT SYSTEMS refueling canal liner and the containment building penetration [10 and does not form part of the containment building pressure boundary. The equipment hatch and air lock doors are fitted with double seals with an interspace test connection. Clamps are provided for restraining the air locli% ors when the se al-int-erspacear-moet air lock >^is pressurized. When multiple openings of the contain-ment air locks occur, the air locks will be tested at least once every 3 days. The air lock chamber and the equipment hatch interseal are tested at 49.2 psig and the airlock door seal 10 interspace.is tested at 14.5 1 0.5 psig. Electrical and mechanical penetrations on the air lock are provided with double seals and test connections. Type B tests are conducted at containment peak accident pres-sure (Pa) as defined in table 6.2.6-1. The acceptance cri-teria and leakage rate limits are given in the facility test 10 specifications. Test methods are described in section 6.2.6.3 below. 6.2.6.3 Containment Isolation Valve Leakace Rate Tests Containment isolation valves are Type C tested in accordance with 10CFR50, Appendix J, as listed in table 6.2.6-3. The process piping, instrumentation tubing, and personnel access penetrations are listed in table 6.2.4-1. Figure 6.2.4-1 shows the location of all test vent and drain connections and the direction in which the isolation valves will be tested. The containment isolation valves for each piping penetration are { tabulated in table 6.2.6-3, together with test method and test direction. The table also indicates the status of the valves i during containment building, Type A test, and whether the sys-tem will be vented to the containment and drained during the Type A test. Type C (and B) tests are performed by local pressurization utilizing either the pressure decay or flowmeter method. For December 1982 6.2.6-21 Amendment 10
? (144) peserietion This change installed two new radio base stations, battery bank, antenna ~ tower, and associated cabling to improve security radio coverage. This change was implemented in Units 2 and 3 during this reporting period and-affected drawing 13-E-NNA-002, which is incorporated into the FSAR by reference. gummary of Safety Evaluation This change did not introduce an unreviewed safety question. The change has no effect on any safety related systems. (145) Description The following improvements were made to the_ security radio system: (1) added a signal quality selection _ system, (2) removed secondary aquelch. capability, (3) installed four-wire audio circuit, (4) improved receiver preamplifier specifications, (5) installed a telephone in the main steam support structure, (6) provided human factors improvements for the central and secondary alarm stations, (7) reduced unauthorized to radio consoles, and (8) extended radio coverage by additional access antenna installations. Those changes vere implemented in all three units during this reporting period. Drawings 13-E-ZAC-004, 13-E-ZAC-015, 13-E-ZAC-016, 13-E-ZAC 017, and 13-E-ZAC-018, which are incorporated into the FSAR by reference, were affected by this change. Summary of Safety Evaluation l This change did not introduce an unreviewed safety question. The equipment in question is not important to safety nor does it affect the operation of a safety system. In addition, the system is not identified as a safe shutdown system in the spurious actuation study, and is not used by the operators to safely shutdown the plant. The cable separation study iJ also not affected. Additions to combustible loadings are negligible and do not appreciably affect the current fire hazards analysis. )( (146) Description This change removed insulation from the pressurizer relief valves and from the common header to the reactor drain tank (RDT). The change was implemented in Unit 2 during this reporting period. Drawings l 13 M-CHP-003 (FSAR Section 9.3) and 13 M-RCP-001 (FSAR Section 5.1) were affected. Summarv of Safeev Evaluation This change did not introduce an unreviewed safety question. The change was made to enhance pressurizer relief valve reliability in the event of possible infrequent weepage. The line in question has little or no effect on containment heat loads and the insulation has no safety impact. 45-
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- 10. Will the change impact the Finnt 'Icch Specs 2. If yes, what ocas,and what'.w/ 3;,. -,' :
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- 12. What.is the impact on the plant 11.the. change is not rada pn Unit. liunt.12, the./7. '
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L d L P4.- 1 l (147) Description This change distributed annunciator window #7A10A inputs to windows
- 7A09A, #7A09B, #7A10A, and #7A11A.
The change was implemented in Units 1 and 2 during this reporting period. Drawings 13-J-HCL-001, 13-J-HCL-002, 13-J-HCL-003, and 13-J-HCL-004, which are incorporated into the FSAR by reference, were affected. Summary of Safety Evaluation l This change did not introduce an unreviewed safety question. The operation and function of the system will remain the same. (148) Description This change installed a Typernette 340 printer in place of the alarm typer for the plant computer, and installed the' demand typer in a new location. The change was implemented in Unit 3 during this reporting periad. Drawings 13 -E-ZJ C- 006, 13 -E-ZJ C-007 and 13-E-ZJC 009, which are incorporated into the FSAR by reference, were affected. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. The chan5e is consistent with the system design bases, and does not i increase the failure probability of the Plant Monitoring System. The equipment is not safety related. (149) Description For the Blowdown Demineralized System, the following changes were made: (1) blowdown control valves were modified, (2) conical orifices installed, (3) the flash tank pressure controller was modified and were (4) balancing globe valves and flow indicators were installed on the Blowdown Heat Exchanger cooling water line. These changes were implemented in Unit 3 during this reporting period and affected drawing 13-M-SCP-004, which is incorporated into the FSAR by reference. Summary of Safetv Evaluation This change did not introduce an unreviewed safety question. The l modifications do not change the operation of the system. No safety related components are affected. ( (150) Description This change involved installation of a spectacle flange and test connection on shutdown cooling relief valves. This affected FSAR Figure 6.3-1 and drawing 13-M-SIP-002 (FSAR Section 6.3). The change was implemented in Unit 3 during this reporting period. 0 l l _
( ,I Summarv of Safety Evaluation This chanSe did not introduce an unreviewed safety question. The i function and design of the valves remains the same. (151) Description l This change moved the power feeds to the Radiation Monitoring System (RMS) control room cabinets J-SQA/B-Col and J-SQA/B-COS from distribution panels D31 and D32 to distribution panels D25 and D26. In
- addition, larger secondary cables were provided from E-PNA-V25 and E-PNB-V26 to distribution panels D25 and D26.
This change was implemented in Unit 3 during this reporting period. Drawings 13-E-PNA-001, 13-E-PNA 002, 13-E-PHA-001, 13-E-PHA-002, 13-E-ZJC-006, 13-E-ZJC-007, 13 - E-ZJ C- 009, 13-E-ZJC-037, 13-E-ZJC-038, 13-E-ZJC-052 i and 13-E-ZJC 053, which are incorporated into the FSAR by reference, were affected. Summary of Safety Evaluation This change did not introduce an unreviewed safety question. Provi ~.ag more reliable power to the RMS cabinets will ensure that communications with the PMS microcomputer will not be interrupted. There is no impact on equipment important to safety. (152) Description This change added backup power to the Post Accident Radiation Monitors J-SQN-RU-139,
- 140, 141, 142, 143 and 144 to be fed from Train B, which is diesel backed up.
This change was implemented in all three units during this reporting period. FSAR Tables 1.8-1, 8.3-3 and 11.5, and drawings 13-E-ZAC-016 and 13-E-PHA-006, which are incorporated into the FSAR by reference, were affected. summat,r of Safetv Evaluation This change did not introduce an unrevived safety question. The modification does not af feet important to safety equipment and will enhance compliance, with technical specifications. l (153) Description This change installed a condensate pot on the drain line from the H /02 analyzer to prevent leakage of gases into the Radwaste Drain 2 System. The change was implemented in Units 1 and 2 during the reporting period and affected drawing 13-N-SSP-001 (FSAP, Section 9.3). ) l Summarv of Safetv Evaluation J l This change did not introduce an unreviewed safety question. The change enhances the operability of the system. l 47-I l
m t -l OFFICIAL WORKING COPY 4*oeoMU .t q ~ ~j i \\ s 10CFESO.59 Pete \\ of 2 ,SAFEIT REVIEW /SAFETI EVALULTION, No. Ece. *]~ D ei/. O Initiating Document 'DC P t o$ T 'Onf L MLT '45 -oc.f. Description of Proposed Change: One,,a d ' Oc Aio E m w e_ Ae em ts Fo m ti.e Ib.s.r e ves Irct e.%, \\4 3,. Ae Wie., .r.... fe 'T's b._.o ['.m...:.e e, - e n G v %-- -J b <. n. e 240* d. Dee,. t r T' 'TJe O v-c. me - EACH QUESTION BELOV SHALL BE ANS**EEED ON ATTACHED RESPCNSE JUSTIFICATION FORK (S) YES No 1. Is a change to the facility as described in the Final Safety Analysis Eeport (FSAE) involved! N/ I 2. Is a change to procedures as described izi the Final Safety Analysis Report (FSAE) involved! 3. Are tests or experiments not described in the Fin ~al ,,: = : y.Acalysis Report (FSAE) to be performed! s/ .Safet ,,2...:.._... 4.;:.:- -p.; : 1st O If any answer to 1 through 3 is "Yes", an FSAE \\ Change Notice (SARCN) may be required.'.,,,; 5,. 1.. (51F.CN O kl A ) .Q ~1*-;;:&~?fkWWf.&W'59:. Q :5 ;???$' M ': .9,' r. b. 4. Does the proposed. change require a change to the \\. Technical Specifications!. - v!. pf . - J.1,O If ansver'to 4 is "Ies". Technical Specification change request and NEC, approval required prior to implementation. D NEC approval obtained by .L I A Date n.mef p--<I- . * (ref erence docu=ent) i -:*.;.; q +: g........... y. 5. Vill the probability of occurrence or the ecosequences of an accident or malfunction of equi; ent icportant to safety i i,,, previously evalueited in t'he FSAE be increased! , /' - g f. . ;.;.;;+,t.q. 6. Vill the possibility for en accident er =alfunction of a differ- ~ ent type than any evaluated previously in the FSAE be created! / ' 7..., V111'the chgin of safety as defined in the 5 asis for a::y t j l - -. Jechnical. Specification be reduced!.. -. 2,. 1,.. , .} ' 4 4.-. M If all ansvers to 4 through 7 are "No". NF.C app oval not v. required prior to change imple=entation. .f ,s 9 r - "r-- If any answer to 5 through 7 is *!es", ~en Unreviewed Safety i [..:.dt. tion 2. Ques tion,exis ts. NEC approval required prior to 1:plenentation. r. NEC approval obtained by M t-Date - (ret erence docu=ent) ~ ~ ~ 6 FSLE Chapter 6/ Chapter 15 analyses pote::tially affected. Overall r system perfor:ance shewn to =cet requirements of safety analysis. .-$j
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CILLNCE DESCRIPTION hELE9ET716 AIRSoosTEP, 455Em8g FRoyn T14E. Doul )comcq 85cLATlotl V40 E5 A AND MobiFy BC CM SYSTEM To suPPty 240 < Mt n:Recyty To THE V ~ Atrunro2.3. .n O: IMPACT ON UNIT DURING I PLEMENTATION I3 b65t6d Cit Ar4GG ' cat 3 MOT BE imPLEMENmbWHILE TNesE v4uts 4g 56RVitC. I6 @rJ-E55E Atr1 AL Auk FEEbWATEA p/ATEC FlowA4ni glig gg '~ se) test WRtLG imPLGMENT1A)6 TNII CH4NGe, IAT w tL( KtCK you tu *ro 4 72 Hour nc. nom (set TECH $1'EL LCO 3.7.i. l,otmoa A). Cunoce 5}iout[66 IMPLErnENTED Midb Ad oOTA6E, _ IMPACT ON UNIT FOLLOWING IMPLEMENTATION rit;fl IMPLCmcrJYnTio r) op 774tS UinM6s 71iE VAUC5 v)lK fus)CDo m A, ecs a s na w a occ Darr a,ne m a sEnr. l u
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- No 9.
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PLANT CHANGE PACKAGE eSGE$***, PCP NO. S Pol-A Cs -O M-OO CHECKLIST PAGE 2 of 2 F PLANT CHANGE CHECKLIST (CONT *D) A AFFECTED TO BE ATTACHED k/g YES NO YES NO~ 13. ELECTRICAL SINGLE LINE DRAWIxc ,7np O 9' d g '14. ELEMENTARY DRAWINGS O E D
- 15. CONTROL AND LOGIC DRAWINGS EO E
O 16. INSTRUMENT LOOPS DRAWINGS &O S' O 17. INSTRUMENT INDEX EO O-18. INSTRUMENT CALIBRATION DATA LIST 13-J-ZZI-003 EO 0 19. LINE DESIGNATION LIST 5O 9' O 20. PIPING AND INSTRUMENT DRAWINGS (P&ID'S) O 6 0 21. PIFING SYSTEM STRESS REPORTS O & [W 6-22. ISOMETRIC DRAWINGS &O
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- 25. EQUIPMENT INDEX 26.
AS-BUILT DRAW'INGS E O O (.27. NUCLEAR DRAWINGS (RADIATION ZONE DRAWINCS) O B' O O' 28. CONSTRUCTABILITY WALKDOWN O O E
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- 32. PLANT MODEL CHANCES O
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- 33. EQUIPMENT QUALIFICATION LIST UPDATE O
7 0 W
- 34.. EQUIPMENT CHANGE EVALUATION O E O
Y INSTRUMENT SETPOINT DOCUMENThool-lot 413) ED E O 35.
- 36. WPP/QCI CWO'S (i.e. 251.1)
O I O SPECIFY
- 37. LIST OTHER (IF AFFECTED BUT NO MENTIONED)
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4 PLANT CHANGE PACKAGE
- M P'i" 3 V
PCP NO. 85-o2-SG-o 7 4-o o . Page 5 of.,9 f- ~ .,C. PCP AS-BUILT VERIFICATION FORM a. TO: C I-O' n?E s O F - a t a g t hlf p sl,{ ! A! Ona uvi Huim au op Sw: 67/s . a v.- t' ..,..DATE: 8 FE 3 d5'/ T- . Tl!IS IS 'IO VERIFY TIIE AS-BUILT (PLANT *ID DWG.) CONFIGURATION Ln:.-? :. 5 -
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~Y~ b*& Q ' T,' ' Q i. ... ASSOCIATED DCPf 203. AF-C8(o @w _s. ..u ri
- 1.. I F.* 7 :: N.~ t VT'$.2f; ';T O.TS 9
AFFECTED DRAWINGS AND/OR AMMENDMENTS (NOT SPECIFIED IN DCP): 9 %=e _,, _.,, e, ;.r_. e -+.. c-.. ., ~~.n ::.. ; r- ~.- ..l ff f h. L 2 ;p7;:r e 1 / ~ ~~. '2'. / SYSTEM ENGINEER DATE 4 ~.~:.... : ~' ~1 C.kHO W TO: PROJECT COORDINATOR STA# < (o 9 8 Z t' DATE: 25 (?-E B7 n. wt. e:;..r,..,
- ~'
s.' ,.THIS IS TO VERIFY THE FOLLOWING FOR PCP# 84 -O2-G.O74-CQ ] 2.>, 2 -.e YES N_f,A INIT/DATE py s-.J l 1.) DRAWING DOCUMENT REGISTER (DDR) UPL/TFf' ],[f) 7/p/qq TO CONSTRUCTION COMPETE STATUS. / 2.) EOF, TSC, AND CONTROL ROOM DRAWINGS O O @# 2 23.s1 STICKS UP DATED TO AS-BUILT. r. xm pp UABRM DATE
~., PLANT CHANGE PACKAGE Pc" Forms PCP NO. 84-02-s a.c y4.c gp Yage 6 of $ (
- 0. :10 AL WORKNG COPY PCP SYSTD4 INSERVICE CHECKLIST VERIFICATION OF COMPLETION:
INITIAL /DATE 1. Test results meet test criteria and requirements YES % N/A D #4 /x-zr-n. I i 2. Surveillance and inservice inspection ,f requirements performed YESh N/AO /r#/x-22-n 3. Verify package is construction complete YES% N/AD h4 / 2 2erp / / 4 . As-Built configuration verified (Plant to Dwg.) YES% N/A O /[R$/.7-AP-r,7 5. NCR closeou't complete YES O N/A /2-Rf-f? ' (/ l 6 All technical specifications changes approved jf / by NRC YES O N/Ah,M70/' 2-2 5-?? v s 7. FCR/ Deviations from approved work package closeout YES O N/A D h l/ d -2 1-fy f/ / Y A'/5~~ 9-25-M YSTEM ENGINEER DATE 8. Operations Procedures Revised YES@ N/AD m esrs=a7 9. Design Documents Register updated (DDR) YES S N/A O c2 23 a=s s7 "10. Verify EOF /TSC/ Control Room Drawing stick . updated to As-Built by ABRM YESE N/AO os e et2.sa7 11. Additional com as require )c #pletion information (Identify YES@ N/AO CE_ crass 7 13205, l94~17 hs Nl [=5 5 07 PROJECT COORDINATOR DATE CONCURRED BY 1~ SHIFT UPERVISOR DATE Sw slw M ~ ~ iPPR0vED 8Y Pru w P w T - AceR u i DAp SYSTEM INSERVICE IS COMPLETE AND ACCEITED e8
r. ) PLANT CHANGE PACKAGE ~ ~ PCP Forias PCP NO.,, A6o 2-S G - o 14 o o - Page 7 o r-t L PCP CLOSEOUT RECORD
- e n-L} {.S,lO, M. @. %u COIN A C G, i:N6 vi wiou.-
v~s s VERIFICATION OF, COMPLETION: YES NO N/A
- e -
INITIALS /DATE - 1* DEPARTMENTAL RACT ;.Wr, i-a us ; i u r..mti.- : nr e nev f., red O 0" . 8..' " ! Jets n e"er 7 A. Maintenance ' 'B. Training-OEO ,'.o.e asrase7 C , a.qSpe,1,l}anc,e.;n a veri.icc O. at / DN@ D. Seeurity 0 .E U % L. i"sissa67 u. c_ E. STA ~ OEO c c. es ex-ger 'I. Computer OEO c2 23 mer G. Radiation Protection OOE ts_. rs erms7 -H. Radwaste O M esr.cr. c I. Emergency Planning OOE d2 25 msa7 Q cyg. zg gggy 3. A1 ara K. Fire Protection S0 <->c. 2s e87 C. YES N/A _ INITIALS /DATE 2. PROVISIONS. FOR FSAR. UPDATE-COMPLETE '!fs) b Tt J/L CC# S fEG67 .3. SIGNED COPIES OF NCR's ATTACHED 0E <-az 2s spa 57 4. ANY CHANGES IN SIMULATION /MODEL AND OTHER OE ASSOCIATED SYSTEM INITIATED C2. Z3[C2,37 5. ADDITIONAL COMPLETION INFORMATION (IDENTIFY AS REQUIRED) b O2 23M/3-[ 6. NOTIFY NUCLEAR ENGINEERING TO UPDATE ORIGINAL DESIGN DOCUMENTS MO ce z3s_g7 ~ 7. EE580 CARDS 50 o2. z s; rsr'er C.CN:LKr.:: 47 .g_ _ b.- ~.,OM af. 'zs M67 8. WPP/QCI CIP BUYOFF(s) ~ - - ~ PERFORMED BY: DATE: M b 67 Project Coordinator PLANT CHANGE PACKAGE CLOSEOUT IS COMPLETE L ____U
1 l PLANT CHANGE PACKAGE AM PCP FORMS etce e r e l PCP NO. Sh"O 2-S G-e> ~14-6 0 (
- 0. F CLA; WORK VG COPY ~
PLANT CHANGE PACKAGE CONTENTS SHEET ATTACHED N/A 1. PLA:.T OIANGE REQUEST h 2. DETAILED DESIGN (DCP) E O 3. NUCLEAR SAFETY EVALUATION E O 4. APPLICABLE NRD CORRESPONDENCE 5. ALL PCP IMPACT REVIEW SHEETS b 6. WO/CWO's IMPLEMENTATION DOCUMENT k 7. OTHERS (LIST AND ATTACH ADDITIONAL SHEETS AS REQUIRED) O e COMPLLTED BY: { i / / / PROJECT COORDINATOR DATE 1 l ...?
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- 88. PROCUREMENT ACTIVITY REQUIRED:
RE STM AINY O R CONDITION hTEs[DfNTjrNPPL8CA'LE j MO ogy a 7 TREND BASIS: E VO LUTION ECOPECNAhGE $ COPE CH ANGE REGULATOMY DEE8GN
- 8. R EPORT ABILITY EV A LU ATION:
u.na4 we. POTT NTI A L M E CO V E R A B LE col'T. SUPPLIER NoYlF8ED OF ygg _ID ENTI FY NOTICE NO. w An m ANTT AND CON ro nM ANCE. NU g /. Pgp g,MBER .. ~
- 9. TITLE OF PRODLEM/CH ANGEz FMwRE Or Checcsea Fes.meurta Iscunew hsTo Cow _n-aOnlb.T)saw 9 A. SUPPLEMENTAL 4NFORM ATION:
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- 10. R EV 8SION D ESCRIPTION:
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IDENTirY ALL INCO$tPOR ATED DCP MODtrtCATION SHEETS: Nb ,.,s y,9;( Eir "N l'** ~ (cs RM-h - @ / - 1-4-Es b c
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- 14. 10C FR 5 0.5 9 DISPOF.iTION DOESh DOES NOT CONSTITUTE AN UNREVIEWED S AFETY OUESTION PER ICCF feTy//
THis .;.,f NRC APPROVAL CIS IS NOT REQUIRED PRIOR TO IMPLEMENTATION NRC APPROVAL DOCUMENTED D e, ,t, ]- w G s ssG m a Y uati \\ cavt ej, ._15 INTER DISCIPLINE COORDIN ATIO N. -5,;;94 .. e*
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- 16. APPROV A LS
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- 19. NOTIFIC ATION OF COMPLET10N:
IDENTirY A ND ATT ACH ALL h NO DCP MODIFIC ATION REC'D DCP MODiriC ATION SHEETS Pf /P nOJE CT FitLD t mG 6h t f R/ O AYE S T A NT UP E N o em t t p sC Lat N T Q U A L 5 T
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- 18. TECHNIC A L DESCRIPTION 7SCCPE OP' WORK
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- 16. TECHNIC A LVEScM 6PTION CE-A p, e
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~... y OFFICIAL WORKING COPY = i ,...I..-6 i' 7 =; 3 PVNCS, l@g-Ap-oB C, DESIGN CHANGE B' ASIS p,gg-A F-O S C.. DCP yrtT-AF-c y_ Revisica f) Perform Safety Review'as required by IP-4.40. j ~2. 2. Al>'A/ISI Feriev (Rev. O DCPs and significant revisions only) If the answer to item 2a, b, or e is "Yes," perform and document an ALARA/ISI reriew in accordance with IP-4.3, or provide justification for exception below? td A Does the alteration add or modify facilities, systems, er cceponents s. which wi22 collect, stere, process, or transport rr.dioactive materials? i D YES K NO b. Does the alteratien add oi modify facilities, systees. or cceponents which are located in radiation areas * (>2.5r_R/hr)? This includes added equipment, cable routing, conduit, piping, and tubint relative to radiation areas affected and/cr traversed. D Yr5 po Does this alteration create or r. edify routine maintenance, inspee. c. tien, or survei31ance requirements in a radiation area' (>2.5mR/hr)? O YES D[N3 ' Rad!stien :ene drawings - see TS1.R. Titures 22.3-1 thr:uth 22.3-10 and Iab2e 22.1-1. (Zones 3, 4, and 5 are 2.5 E/hr.) 3. Is a Design Criteria Chante Eequest required? If Jes, identify DCER number and.cceplete Safety Significance Assessment (IP-4.34): D TES T/NO l_ Engsneerant Grcup Supervasor huclear' Creep Supertas q j r. (' . ' ' If the design chante aff ects the design basis or design criteria, iecord I - date on which the A.NFP Litersint Xar.ater was notified and concurred: p.
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7, s ' OFFICIAL WORKING COPY c I PVNCS Coh71CURATION DOCUMENT CEECC.IST m DCP checklist: DCP Nos: _\\d a.a:- o cr-t a m w - o ct_ ?&-Ac-oce Rev. n In addition to the docu=ents listed in the body of the DCP, indicate by cc:pleting this checklist the other impacted docu=ents which vill be revised. Indicate applicable initiating docunent (e.g., DFS, DCN or ECI Affected: List) which co:pletes Document Title / Identification Tes No actien: Licensing Docu=ents O& saaCN No. Design Criteria OE DCza No.
System Description
OE SDza No-DP3 No. Test Guidelines Design Calculations OO Rev No-O6 SCN No-P.*0. Specification Construction Specification C@ SCN No. ASME III Design Specification SCN No. Supplier' Instruction Manuals Og s;cy yo. i l l hping System Stress Reports OE sev. No. Equipment Qualification Reports Og 53:3 po, Instru=ent Data Sheets [g scs yo. General Adange=ent Dravings O@ DCN No. ( Radiation Zone Drawings [@ r;y ye. Tire Hanards C@ DON No. ISI Dravings C@ DO; No. bb TC No. -], Instru=ent I dex___ OO Other l . S. ...}}