ML20151F153
| ML20151F153 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/11/1988 |
| From: | Miltenberger S Public Service Enterprise Group |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NLR-N88056, NUDOCS 8804180109 | |
| Download: ML20151F153 (27) | |
Text
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.s Public Service Electric and Gas Company Steven E. Miltenberger Public Service Electric and Gas Company P.O. Box 236, Hancocks Bridge, NJ 08038 609 339-4199 Vce Presdent -
Nuclear Operations April 11, 1988 NLR-N88056 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:
UPDATED FINAL SAFETY ANALYSIS REPORT, REVISION 0 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas Company (PSE&G) has just completed updating the Hope Creek Final Safety Analysis Report (UPSAR) pursuant to 10 CFR 50.71(e)(3)(i).
Controlled copies will be provided under separate cover in accordance with 10 CFR 50.4(b)(6).
To facilitate your review a brief explanation of each change is provided in Attachment 1.
These changes include incorporation of some 250 question responses, incorporation of pertinent SSER information, revisions due to plant design changes, and revisions addressing license conditions.
It should be noted that the question and response section has been removed from the UFSAR as we consider the pertinent information contained therein to have been appropriately incorporated into the applicable UFSAR section.
Upon your receipt of the Updated Final Safety Analysis Report (UFSAR), the Hope Creek Final Safety Analysis Report (FSAR) is considered superceded, decontrolled and should be disposed of accordingly.
Should you have any questions regarding this submittal, please do not hesitate to contact us.
Sincerely,
//
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\\j Attachment
Document Control Desk 2
04-11-88 C
Mr.
G.
W.
Rivenbark USNRC Licensing Project Manager Mr. G.
W. Meyer USNRC Senior Resident Inspector Mr.
W.
T.
Russell, Administrator USNRC Region I Mr.
D.
M. Scott, Chief Bureau of Nuclear Engineering Department of Environmental Protection 380 Scotch Road Trenton, NJ 08628
e Q:
l ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE F1.2-1 Design Change Package 5-EC-0001 deleted the site F2.4-9 security fences between HCGS and SGS and added F3.4-3, SH 2/4 additional lighting thereby permitting un-controlled access between stations.
This change did not alter the security requirements for access to the HCGS/SGS compl6x at Artificial Island.
1.8.1.52 Incorporation of information previously provided in the rasponse to Question 460.9.
1.8.1.68 Incorporation of information previously provided in the response to Question 640.12.
1.8.1.68.Z Discussion provided which deferred the initial turbine-generator testing to 10% thermal power from 5% in order to assure adequate steam for prossure control.
This change was made in accordance with Facility Operating License NPF-50, License Condition 2.C(ll) and accordingly was previously submitted in a letter f rom C. A.
McNeill, Jr. to T.
E. Murley dated August 25, 1986.
1.8.68.3 Incorporation of information previously provided in the response to Question 640.8.
1.8.1.68.3.b.1 The permissible primary containment instrument 9.3.1.2 air particle size was increased from 3.0 to 50
.9.3.1.4 micrometers based on equipment supplier docu-mentation and engineering evaluations.
These changes were made in accordance with Facility i
Operating License NPF-50, License Condition 2.C(11) and accordingly were previously trans-mitted in a letter f rom C. A. McNeill, Jr. to T.
E. Murley dated July 21, 1986.
l.8.1.75 Incorporation of inf ormation previously provided in the response to Question 421.8.
1.8.1.84 Change the revision level of Regulatory Guide T5.2-2, PG 7/8 1.84 from Revision 18 (8/81) to 24 (6/86) in order to permit the use of ASME Code Case N-411 in the snubber reduction program.
1.8.1.97.5 Incorporation of information contained in SSER 2, Appendix M.
PAGE 1 OF 25
ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 1.8.1.137.
Design Change Package 4-HM-0152 revised the 9.2.4.5 diesel generator fuel oil storage tank level alarm logic.
1.8.1.142 Incorporation of information previously provided in the response to Question 220.26.
1.10.2.I.D.2 Incorporation of information previously provided 1.10.2.II.B.1 in the response to SRAI (1).
1.10.2.II.B.2 1.10.2.II.B.3 1.10.2.II.D.3 1.10.2.II.E.4.1 1.10.2.II.E.4.2 1.10.2.II.K.3.15 1.10.2.II.K.3.16 1.10.2.II.K.3.18 1.10.2.II.K.3.22 1.10.2.II.K.3.24 1.10.2.II.K.3.25 1 10.2.II.K.3.27 1.10.2.II.K.3.28 1.10.2.III.A.l.1 1.10.2.III.A.l.2
.l.10.2.III.D.3.3 1.10.2.II.F.2 Incorporation of information previously provided in the response to Question 421.21.
1.10.2.II.K.3.13 Incorporation of information previously provided in the response to Question 421.12.
1.12.3.7 Incorporation of information previously provided in the response to Question 430.31 and information contained in SSER 4, Section 14.2.
1.14.1.36.2 Revisions reflect the resolution of the BWR 1.14.1.119.2 stability issue (Generic Letter 86-02) precluding the need for future stability analyses.
2.2.3.1.5 Incorporation of information contained in SSER 4, Section 2.2.2.
2.3.2.1.1.1 Incorporation of information previously provided 2.3.2.1.6 in the response to Question 451.6.
l T2.3-39-43 l
l PAGE 2 OF 25 L
ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 2.3.4.3 Incorporation of information previously provided P2.3 in the response to Question 451.16.
Addition-ally, clarifications and corrections are provided regarding plant cross-sectional areas.
^2.3.5.3.3
' Incorporation of inf ormation previously provided in the response to Question 451.18.
2.4.1.2.1 Incorporation of information previously provided in the response to Question 240.5.
2.5.1.2.1 Incorporation of information previously provided
-in the response to Question 231.2.
2.5.1.2.3 Incorporation of information previously provided P2.5-63 in the response to Questions 231.6 and 231.7.
2.5.2.4.1.1 Incorporation of information previously provided T2.5-21-24 in the response to Question 230.8.
P2.5-62 2.5.4.2.2 Incorporation of information previously provided in the response to Question 241.5.
2.5.4.5.1 Incorporation of information previously provided in the response to Question 241.15.
2.5.4.6.3.1 Incorporation of information previously provided P2.5-64,65 in the response to Question 241.17.
2.5.4.10.1.1 Incorporation of information previously provided P2.5-67-96 in the response to Question 241.25.
2.5.4.10.1.2 Incorporation of information previously provided P2.5-66 in the response to Question 241.28.
3.4.1.1 Incorporation of information previously provided T3.4-4 in the responses to Questions 410.3, 410.5l 410.7, and 410.8.
3.4.1.1 Design Change Package 4-EMC-86-Oll added 9.5.1.2.15.1 Vestibule 3301A and the necessary heat detector 9A.6.3 to Entry Corridor 3301 which in turn permitted T9A-1 using Door 3301A in an open position.
Therefore, T9A-4 the state-ment that all exterior doos are T9A-6 normally closed (Section 3.4.1.1) was deleted and information regarding the heat detector was added to Sections 9.5 and 9A.
PAGE 3 0F 25
ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 3.4.1.5 Incorporation of information contained in SSER 4, 3.4.3 Section 3.4.1.
3.5.1.1 Incorporation of information previously provided in the responses to Question 430.77 and 430.78.
3.5.1.3 Incorporation of information contained ir, SSER 6, 3.5.4 Section 3.5.1.3.
3.5.2 Incorporation of information previously provided F3.5-29 in the responses to Questions 410.18 and 410.19.
3.6.1.2.1.3 Incorporation of information previously provided in the response to Question 210.20 and information contained in SSER 3, Section 3.6.2.
3.6.1.2.1.19 Incorporation of information previously provided in the response to Question 430.120.
3.6.1.2.2 Incorporation of information previously provided in the last paragraph of the response to Question 430.82.
Additionally, the minimum flash point of No. 2 diesel fuel oil is 123'F, as identified in ASTM Standard D975-77 rather than 100*F originally shown in the question response.
3.6.2.1.l' Incorporation of information previously provided in the response to Question 210.15 and information contained in SSER 3, Section 3.6.2.1 and Appendix 0 and SSER 5, Section 3.6.2.1.
3.6.2.6.3.2 Incorporation of information previously provided in the response to Question 210.22.
3.6.4 Incorporation of information contained in SSER 3, Sections 3.6.2 and 3.6.2.1 and SSER Appendix O.
3.8.2.1.5 Incorporation of information previously provided T3.8-20 in the response to Question 210.37.
3.9.1.4 Incorporation of information previously provided in the responses to Questions 210.27 and 210.34.
3.9.1.4.13 Incorporation of information previously provided in the response to Question 210.36.
PAGE 4 OF 25
ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 3.9.2.3.2.8 Incorporation of information previously provided 3.9.2.3.2.10
'in the response to Question 271.3.
3.9.2.3.2.16 3.9.3.2.7.1 Incorporation of information previously provided in the response to Question 271.4.
3.9.3.3.1 Incorporation of information previously provided T3.9-31 in the response to Question 210.45.
F3.9-7-16 3.9.3.4 Incorporation of information previously provided 3.9.3.4.1 in the responses to Questions 210.46, 210.47 and 3.9.3.4.5 210.50 and information contained in SSER 4, 3.9.3.4.6 Section 3.9.6.
3.9.7 3.9.6.1 Incorporation of information contained in SSER 4, 3.9.6.2 Section 3.9.6 and SSER 5, Section 3.9.6.
T3.10-3 Design Change Package /DCP 4-HCJ-86-1065 replaced the main steam line/offgas pretreatment log radiation monitors with comparible monitors but more stable, reliable and easy to calibrate.
3.11.1.1.1 Reference to and information within Tables 3.11.1.2 3.11-4, 5& 6, Mechanical, Electrical and Safety-3.11.2.6 Related Equipment in Harsh Environments, was T3.ll-3 deleted in favor of that provided in the Environ-mental Qualification Summary Report (EOSR) provided under separate cover.
3.11.1.4 Incorporation of information previously provided in the response to Question 410.100.
3.11.2.1 Incorporation of information contained in SSER 5, 3.11.7 Section 3.11.2.
T3.ll-la, PG 6/9 Changes to various room temperatures are T3.ll-lc, PG 4,5, necessary to eliminate discrepancies between 11, 12/21 design documents and FSAR tables.
The revisions reflect as-built conditions.
4.6.3.1.6 Clarification of the 5-year maintenance life test program wording for CRDS to eliminate incorrect interpretations.
t PAGE 5 OF 25 I
.~
1 ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 5.2.2.2.2.1 Incorporation of information previously provided F5.2-14 in the response to Question 440.6.
5.2.2.4.2.1 Incorporation of information previously provided
~
S.2.6 in the response to Question 440.7.
Additionally,.
the discussion regarding SRV performance testing following refueling activities has been revised to reflect Tech. Spec. surveillance raquirements.
~5.2.3.2.2 Design Change Package 4-ECM-86-112, Rev. 1 T5.2-8 provided a zine injection system to inject zine T9.5-3 oxide into the reactor feedwater and maintain a zinc concentration of 5 to 10 ppb.
This change inhibits stainless steel corrosion which reduces Co-60 buildup and therefore reduces radiation levels by a factor of three or more over the course of the plant's life.
This modification follows the guidelines of EPRI Report NP-4072, is operated consistent with BWR's with brass condensor/powdex demineralizer combinations and meets the recommendations of General Electric.
5.4.6.2.1.2.d Design Change Package-4-HCE-86-727 delays the 5.4.6.2.2.2.i.1 initiation of the ramp signal generator until MOV E51-HV-F045 begins to reopen following a 15-second time-delay.
This change permits an adjustable partial valve opening to aid in controlling the initial acceleration of the Reactor Core Isolation Cooling (RCIC) system turbine, effectively allowing valve F045 and the RCIC governor valve to act as throttle valves for the RCIC turbine.
T5.2-2 Incorporation of information previously provided in the response to Question 210.12, 5.4.6.1.1.1 Incorporation of information previously provided in the response to Question 440.10.
5.4.6.2.1.2.e,f,g,m Various editorial changes to correct valve 5.4.6.2.5.1.r designation errors including test loop operation 5.4.6.2.5.2.j inaccuracies.
5.4.6.2.2.2 Incorporation of information previously provided in the response to Question 440.13.
1 PAGE 6 OF 25 1
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ATTACHMBNT 1 SECTION
SUMMARY
OF CHANGE 5.4.6.2.4 Incorporation of information previously provided in the response to Question 440.12.
5.4.7.2.6.a-Discussion regarding the method used to. flush and prewarm the RHR piping prior to establishing shutdown cooling was revised to reflect Operating Procedure OP-SO.BC-01.
Appendix'5A Incorporation of information previously provided T5A-19-26 in the responses to Questions 251.1, 251.2, 251.4, 251.6, 251.7.
6.2.1.1.3.6 Incorporation of information contained in SSER 3, Section 6.2.1.7 and information previously provided in the responses to Questions 480.3 and 480.4.
6.2.2.2.2 Incorporation of information contained in SSER 6, Appendix O.
6.2.3.2.3 Incorporation of information previously provided 6.2.3.2.4 in the responses to Questions 480.12 and 480.14.
F6.2-52-58 6.2.4.1.2 Incorporation of information previously provided in the response to Question A50.2.
6.2.4.3.2.5 Changes discuss the reclassification of the TIP 6.2.4.3.2.6 system containment penetrations under GDC 56 T6.2-16 PG 4,5/33 rather than GDC 54 pursuant to IE Information Notice 86-75.
6.2.4.3.2.22 Incorporation of information previously provided in the response to Question 480.23.
6.2.4.4.7.a Revised incorrect description of Appendix J test T6.2-24 PG 2,3/17 requirements for the drywell chilled water pressure safety valves.
6.2.4.5.4 Incorporation of information previously provided in the response to Question 480.23.
6.2.6.1 Incorporation of information previously provided 6.2.6.3 in the response to Question 480.31.
I 1
PAGE 7 OF 25
ATTACHMENT 1 SECTION
SUMMARY
OF CHANCE o
. 6.2.7.2 Incorporation of information contained in SSER 2, Section 6.2.7.
T6.2-6.
The peak suppression pool tenperature during a recirculation line break or main steam line break (i.e. Case E was increased from 209*F to 212*F to correct an editorial error.
T6.2-24, PG 16/17 Incorporation of information previously provided in the response to Question 480.34.
6.3.2.2.1 Incor?nration of information previously provided in t' responses to Questions 440.22 and 440.23.
6.3.2.2.3 Incorporation of information previously provided in the responses to Questions 440.24 and 440.26.
6.3.6 Incorporation of information previously provided in the response to Question 440.27.
6.8.1.1 Revised discussion of the reactor building 6.8.1.2 relative humidity t.o address maximum calculated 6.8.2.1 rather than 100% RH and to delete references to 6.8.2.2 specific temperature rises in "RVS process flow.
These changes reflect as-built plant conditions while still maintaining FRVS design basis charce.21 absorber RH.
7.1.2 Incorporation of information previously provided 7.1.2.1 in the responses to Questions 421.5, 421.11 and F7.1-6-10 421.23.
7.1.2.3 Incorporation of informacion previously provided in the response to Question 421.22.
7.1.2.4 Incorporation of information previously provided i
in the response to Question 421.7.
7.1.2.4 Incorporation of information contained in SSER 5, Section 7.2.2.4.
7.1.2.5.1 Incorporation of information previously provided 7.1.2.5.2 in the responses to Questions 421.10 and 421.13 T7.1-4 and contained in SSER 5, Section 7.2.2.7.
t 5
I PAGE 8 OF 25 1
1
i ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 7.1.2.7 Incorporation of information previously provided in the responses to Questions 421.2 and 421.3.
7.1.2.9 Incorporation of information previously provided in the response to Question:421.6 and contained in SSER 5, Section_7.3.2.5.
7.2.1.1 Incorporation of information previously provided F7.2-2 in the responses to Questions 421.16 and 421.20.
7.2.1.3.6 Incorporation of information previously provided in the response to Question 421.17.
7.2.1.3.7 Incorporation of information previously provided in the response to Question 421.19.
7.3.1 Incorporation of information contained in SSER 5, Section 7.3.2.6.
7.3.1.1.1.2 Incorporation o'f information previously provided in the response to Question 421.31.
7.3.1.1.2 Incorporation of information previously provided 7.6.1.3.2 in the response to Question 421.33.
7.3.1.1.ll.l.b.l.d.
Design Change Package-4-ECM-86-696 deleted the 9.2.1.4.1 cyclone separators from the station service water system to assure adequate lubricating water flow to station service water pumps and improve reliability of the head tank lubricating water system.
7.3.2.1.3 Incorporation of information previously provided in the response to Question 421.35.
7.3.2.2.4 Incorporation of information contained in SSER 4, Section 7.2.2.3.
7.4.1.1.2 Incorporation of information previously provided in the response to Question 421.39.
7.4.1.2.2 Incorporation of information previously provided in the response to Question 421.40.
T7.4-3 PG 8/8 Revise descriptive remarks for the control area chilled water system to describe the as-built method of remote starting the AK400 and AK403 chillers-PACE 9 OF 25
ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 7.5.2.6 Incorporation of information previously provided 7.5.3 in the response to Question 421.42 and information contained in SSER 3, Section 7.4.2.1.
T7.5-1, SH 2/23 Provide discussion of the readout range on the instrument panel monitoring euppression pool water temperature.
This change was provided for clarification only, does not reflect any plant changes and is consistent with Section 6.2.1.1 and 9.4.5.1 as well as Tech. Spec. 3.6.2.1.
T7.5-1, SH 2,3,23/23 Design Change Package-4-ECE-86-964 provided Class lE backed power for the neutron monitoring system SRM/IRM drive motors.
The appropriate f ootnote was provided to clarify this design.
T7.5-1, SH 13,16/23 Correct the power supply identified for the condensor vacuum and high radioactivity liquid tank level display parameter from UPS to offsite power.
This change 13 editorial and involved no physical plant modifications.
7.6.1.2.2 Incorporation of information contained in SSER 4, Section 7.6.2.1.
7.6.1.4.3 Incorporation of information previously provided in the response to Question 421.47.
7.6.2.2.4 Incorporation of information previously provided-F7.6-12 in the responses to Questions 421.3 and 440.21.
7.6.2.5.1 Incorporation of information contained in SSER 3, Section 7.6.2.4.
7.7.1.1.2.1 Incorporation of information previously provided in the response to Question 421.56.
7.7.1.6.3.5 Design Change MDCP 4-HM-0010 adds isolation valves and a
'T' test tap in the RPS pressure switch instrument tubing to reduce time spent in a contaminated area during testing and EHC oil npillage during calibration.
PAGE 10 OF 25 r
.s ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 7.7.1.6.5.1 References to control room annunciation for the 7.7.1.6.5.2 pressure _ regulator turbine-generator system (PRTGS) pressure regulator failure, and control roem control capability f or preesure regulator selection were deleted.
These changes reference as-built configuration in light-of the avail-ability of such functions on the EHC control cabinet in the auxiliary control rocm.
7.7.2 Incorporation of information nceviously provided in the responses to Questions 4 21.48, 421.50, 421.51, 421.53,.421.54 and information contained in SSER 3, Section 7.7.2.1.
8.1.4.6 Incorporation of information previously provided in the response to Question 430.11.
8.1.4.9 Incorporation of information previously provided
'F8.1-1 in the response to Question 430.41.
8.1.4.10 Incorporation of information previously provided in the response to Question 430.42.
8.1.4.12 Incorporation of information previously provided in the response to Question 430.45.
8.1.4.14.1.3 Incorporation of information previously provided T8.1-2 in the response to Question 430.33.
F8.1-2 8.1.4.22 Incorporation of information previously provided in the response to Question 430.23 and slightly reworded to clarify statements.
8.1.7.4 Incorporation of information previously provided in the response to Question 430.56.
8.2.2 Incorporation of information previously provioed in the response to Question 430.8.
8.3.1.1.3 Incorporation of information previously provided in the response to Queetion 430.58.
i 8.3.1.1.2.7 Incorporation of information previously provided in the response to Question 430.19.
PAGE 11 OF 25
ATTACHMENT-1
,$dCTION
SUMMARY
OF CHANGE 8.3.1.1.3.7 Information regarding the results of the pre-operational test of the standby di.esel generators was added.
This information also provides the final response to Question 430.17 which has been deleted in favor of this section.
8.3.1.1.3.10 Incorporation of information previously provided
.n t.se responses to Questions 430.14 and 430.62.
8.3.1.1.5 Incorporation of information previously provided in the response to Question 430.20.
8.3.1.5.2 Incorporation of information previously provided in the response to Question 421.25.
8.3.2.1.2.1 Incorporation of information previously provided in the response to Question 430.25.
T8.3-8, PG 1/2 Design Change Package 4-ECM-86-355 revised the motor size for the HPCI condensor cooling water supply valve from 0.15 HP to 0.361 HP to satisfy original design requirements.
9.1.1.2 Incorporation of information previously provided 9.1.2.2.2.2 in the responses to Questions 410.36 and 281.13.
9.1.2.3.3 Incorporation of information previously provided in the response to Question 410.42.
9.1.3.3 Incorporation of information previously provided in the response to Question 410.56.
9.1.3.4 Incorporation of information previously provided in the response to Question 410.53.
9.1.3.6 Incorporation of information previously provided in the responses to Questions 410.46 and 410.47.
9.1.4.1 Incorporation of information previously provided in the response to Question 410.59.
9.1.4.2.10.1 Incorporation of information previously provided in the response to Question 410.37.
9.1.4.3 Incorporation of information previously provided in the responses to Questions 410.60 and 410.61.
PAGE 12 OP 25
2 ATTACHMENT 1
-SECTION.
SUMMARY
OF CHANGE 9.1.5.3 Incorporation of information previously provided in the response to Question 410.45.
9.1.5.3.1 Incorporation of inf ormation previously provided
,in the response to Question 410.62.
T9.1-23, PG 2&3/3 Addition of a cross-reference to FSAR Section 9.1.5.3.2 f or the refueling bellows guard - ring.
9.2.1.4.3.1 Incorporation of information previously provided in the response to Question 410.73.
9.2.1.4.3.2 Incorporation of information previcusly provided in the responses to Questions 410.65 and 410.66.
9.2.1.5 Incorporation of inf ormation previously provided in the response to Questions 410.64 and 410.68.
9.2.2.2 Incorporation of information previously provi.ded in the response to Question 410.79.
9.2.3.2 Incorporation of information previously orovided in the rec,onse to Question 410.82.
9.2.4.2 Design Change Package 7065 (DCR-HCuKD-003) revised the Potable and Sanitary Water Jystems (PSWS) by chlorinating the fresh water supply at the wellhead rather than upstream of the firewater storage tanks and changing the system flow path to bypass the firewater storage tanks thereby separating the aystem from the water supply for fire protection.
9.2.4.2 Incorporation of information previously provided in the response to Question 410.84.
9.2.6.5.1 Various editorial revisions to correct 9.4.6.2.d discrepancies with design documents, drawings and 9.4.6.3 specifications as well as as-built plant 9.5.'.2.1 configuration.
9.5.2.2.4
-9.2.8.2 Design Change Package 4-HMM-86-1282 rerouted RACS piping negating the requirement to isolate radwaste and open reactor building isolation valves to provide cooling to the emergency instrument air compressors during a LOP.
Appropriate deletions made.
PAGE 13 OF 25 k
AE
.4 ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 9.2.8.2 Design Change Package 4-HMJ-86-1256. removed the process start inhibit signal (PSIS) from the RACS i
pump motor start logic and an interlock from the i
"B" RACS pump so that both RACS pumps will automatically start after a LOP.
T9.2-4, SH 2/2 Correct errors in the summation of SACS total flow rates and heat loads.
Actual values within table and system remain unchanged.
T9.2-8 The description of the auxiliary building safety related, panel room chilled water system chiller capacity was reduced from 200 to 180 tons.
This change reflecte as-built conditions with chilled water entering at 55'F and exiting at 45*F under a flowrate of 430 gpm.
This revision is for consistency, there was no physical change to the system nor its operation.
9.3.1.2 Incorporation of information previously provided 9.3.1.4 in the responses to Questions 410.87 and 410.88.
9.3.3.2 Incorporation of information previously provided in the response to Question 410.90.
T9.3-4, PG. 2/2 Design Change Package 4-BMM-86-358 (DCP-7212) changed the oily waste drain pump impellers to provide sufficient pumping head to pump the waste oil stream from the sump to the separators.
Changes made to reflect as-built and revise pump parameters.
9.4.1.1 Incorporation of information previously provided 9.4.6.1 in the response to Question 451.5.
9.4.7.1 T9.4-22 9,t.1.3 Incorporation of information previously provided in response to Question 410.95 9.4.2.1.d Design change MDCP 4-HM-0058 increased the 9.4.2.2.6 reactor and turbine building air temperature controllers from 40*F to 50*F to prevent freezing of HVAC cooling coils.
i PAGE 14 0F 25 I
s ATTACHMENT 1 SECTION
SUMMARY
OF CHANGE 9.4.3.1.2.2 Incorporation of information previously provided 9.4.3.2.1.2 in the responses to Questions 410.104 and 9.4.3.2.1.3 410.106.
9.4.5.1.a Provide description of localized hot spots in the drywell in accordance with Safety Svaluation H-1-GTXX-MSE-0669-0 9.4.6.1 Incorporation of information previously provided in the response to Question 410.110.
9.4.6.3 Incorporation of information previously provided 9.5.8.2 in the re:'ponse to Question 430.145.
F9.5-47 T9.4-12, PG 1/7 Design Change Package 4-HMO-86-0186 evised the turbine building supply air flow re na 180,000 to 172,000 cfm to maintain Juilding at a slight negative pressure in acev. dane with 9.4.4.1.e.
9.5.1.2.3.2 Design change DCR 4-RMM-86-455 eliminated the diesel fire pump automatic test timer to prevent inadvertent starts and utilized Procedure M10-SHT-009 instead.
9.5.1.2.24 Incorporation of information contained in SSER 5, Section 9.5.1.6.
9.5.1.2.27 Incorporation of information previously provided in the response to Question 430.86.
9.5.1.4.2 Additional discussion provided on the operational testing and inspection requirements for various fire protection systems.
This information was previously provided as Enclosure 1 to the letter from C. A. McNeill, Jr. to E. Adensam dated May 13, 1986 in support of the deletion of the fire protection program elements from the Technical Specifications.
This change was approved in SSER 6, Section 9.5.1. and Table 16.1, 9.5.1.6.3 Incorporation of information contained in SSER 5, 9.5.1.6.8 Section 9.5.1.4.
9.5.1.6.10 i
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OF CHANGE 9.5.1.6.20C Incorporation of information previously provided in the response to Cuestion 280.4.
9.5.2.1 Incorporation of information previously r ovided in the response to Question 410.65.
9.5.2.2.l' Incorporation of information previously provided in the response to Question 430.65.
9.5.2.2.4 Incorporation of information previously provided in response to Question 430.67.
9 5.2.3 Incorporation of information previously provided r
in response to Question 430.67.
9.5.3.2.2 Incorporation of information previously provided in the responses to Questions 430.72 and 4 30.75.
9.5.3.3 Incorporation of information previously provided in the response to Question 430.74.
9.5.4.2.1 Incorporation of information previously provided in responses to Questions 430.87 and 430.79.
Additionally, the restriction associated with 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> settlement time following a fuel oil delivery has been clarified by specifying the tar.k level at which this restriction applies.
9.5.4.2.2 Incorporation of information previously provided F9.5-21 in the response to Question 430.92.
9.5.4.2.6 Incorporation of information previously provided F9.5-47 in the responses to Questions 430.80, 430.89 430.93, 430.94 and 430.95 and information contained in SSER 1, Section 9.5.4.2.
9.5.4.3 Revise the type of filter in the standby diesel generator fuel oil fill connection from duplex to simplex to reflect as-built conditions.
This change correct an inconsistency with previously revised Table 9.5-4, Sheet 2 of 2.
9.5.4.3 Incorporation of information previously provided in the response to Question 430.88 and information contained in SSER 1, Section 9.5.4.2.
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OF CHANGE 9.5.4.5 Incorporation of information previously provided T9.5-20, 21 in the response to Question'430.76.
9.5.5.2.1 Incorporacion of information previously provided in the responses to Questions 430.106, 430.108, 430.142 and 430.113.
9.5.5.3 Incorporation of information previously provided in the response to Question 430.103.
9.5.5.5 Incorporation of information previously provided T9.5-22, 23 in the response to Question 430.104.
'9.5.6.2 Incorporation of information previously provided F9.5-46 in the responses to Questions 430.64, 430.114 and 430.122, 9.5.6.2 Design Change Package 4-HM-0265 revises the 9.5.6.5 diesel gen 9rator starting cir low pressure alarm T9.5-9 in accordance with Facility Operating License NPF-57, Ameadment 12 9.5.6.3 Incorporat:on of information previously provided in the response to Question 430.124.
9.5.6.4 Incorporation of information previously provided in the response to Question 430.117.
9.5.6.5 Incorporation of information previously provided
- T9.5-24, 25 in the response to Question 430.115.
9.5.7.2 Incorporation of information previously provided F9.5-48 in the responses to Questions 430.135, 430.136 a n.d 130.139.
9.5.7.2.1 Incorporation of information previously provided in the re.ponses to Questions 430.131, 430.132 and 430.~.13 9,5.7.2.2 Incorp3 ration of information previously provided in the responses to Questions 430.125 and 430.126.
9.5.7.2.3 Incorporation of information previously provided in the response to Question 430.137.
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OF CHANGE 9.5.7.3 Incorporation of information previously provided in the response to Question 430.138.
9.5.7.4 Incorporation of information previously provided
.in the response to Question 430.125.
9.5.7.5 Incorporation of information previously provided
-T9.5-26, 27 in the~ responses to Questions 430.125 and 430.127.
9.5.8.2 Incorporation of information previously provided in the response to Question 430.143.
J9.5.8.3' Incorporation of information previously provided-in the responses to Questions 4 30.142 and 430.146 9.5.8.5 Incorporation of informa*. ion previously provided T9.5-28, 29 in the response to Question 430.140.
T9.5-3 Information providing regarding the addition of four propane tanks in the yard adjacent to the circulating water pump structure.
9A.l.8 Addition of a description of the limitations for 9A.4.2.c transient combustible loadings.
This change reflects stetion administrative procedures and ti c Dire Hazards Analysis.
9A.6.5.1.e Incorporation of information previously provided in the letter f rom C. A. McNeill, Jr. to E.'Adensam dated December 3, 1985.
T9A-2, PGS. 12, 13/14 Design Change Package 4-HME-86-1095 corrected T9A-3, PG. 9/16 documen'tational discrepancies in valve system nomenclature listed with these tables - no functional changes resulted.
10.2.2.5.1 Incorporation of information previously provided in the response to Question 430.155.
10.2.2.6 Incorporation of information previously provided in the response to Question 430.151.
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10.2.3.6 The Inservice Inspection Program description for the turbine system was revised to reflect General Electric recommendations and missile generation probability, thereby satisfying Facility Operating License NPF-50, License Condition 2. C (-3 ).
This information was previously transmitted in a letter f rom C. A. McNeill, Jr.
to E. Adendam on July 7, 1986.
10.2.3.6 Incorporation of information previously provided in the response to Question 430.154.
10.3.2 Incorporation of information previously provided in the responses to Questions 410.113 and 410.114.
10.4.1.3.1 Incorporation of information previously provided in the response to Question 430.161.
10.4.1.3.3 Incorporation of information previously provided in the response to Question 410.115.
10.4.1.4 Incorporation of information previously provided in the response to Question 430.469.
10.4.2.2 Incorporation of information previously provided in the respan 3 to Question 430.160.
10.4.4.3 Incorporation of ir, formation previously provided in tbs response to Question 430.169.
I 10.4.4.3.?.
Incorporation of information previously provided in the response to Question 430.168.
10.4.4.4 Incorporation of information previously provided in the response to Question 430.167.
10.4.5.2 Design Changes DCP-381 and DCP-391 modify the circulating water system (CWS) by replacing s alfuric acid with sodium hydroxide to prevent l
ccaling.
10.4.6.2.1 Incorporation of information previously provided in the response to Question 281.10.
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OF CHANGE 10.4.6.4 Incorporation of information previously provided in the response to Question 281.9.
10.4.7.2.2 Incorporation of information previously provided in the response to Question 410.116.
T11.2-10 Revised the detergent drain filter description to reference replaceable cartridges rather than specifying various ratings.
This change provides operational flexibility and is consistent with the design requirements of the liquid waste-management system.
11.3.2.1.1 The hydrogen concentration value downstream of the offgas cooler condenser was increased from 0.1% to 1.5% by volume.
This increase reflects as-built hydrogen recombiner performance and cryogenic offgas system decign throughout all plant power levels and is consistent with alarm setpoints and allowable values in Technical Specification 3.11.2.6.
T11.4-3 Design Change Package 4-HC-0109 replaced the centrifuge metering pump with a 2 HP rather than 1 HP motor in order to permit operation in accordance with design.
11.5.2.2.5 Design Change Package 4-HCJ-86-837 included additional int 6 clocks for the Liquid Radwaste Radiation Monitoring System isolation and alarm functions.
The description provides clarification for the use of high radiation, high 1
l radwaste flow, low blowdown or sample flow, and
~
equipment failure.in isolating the liquid radwaste discharge.
Tll.'J-l Changes to the Radiation Monitoring Systems instrumentation ranges and minimum detectable concentrations reflect as-built plant conditions as verified by vendor calibration.
No physical plant change was involved.
12.3.2.2.3 Incorporation of informatiori previously provided in the response to Question 471.4 r
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13.1 Entire section revised to reflect current organizational structure of the Nuclear Department.
13.2.1.1.2 Incorporation of information previously provided in the response to Question 630.10 and slightly revised to reflect current course content elements.
l 13.2.1.1.3.2 Incorporation of information previously provided 13.2.1.1.5.1 in the response to Question 630.4.
13.2.1.1.5.2
-13.4.1.4 Revisions made to reflect actual requirements of 13.4.2.1 Technical Specificationn 6.5.1.6.f,g and h, 13.4.2.1.3.b 6.5.2.4.1 and 6.5.2.4.4.b respectively.
14.2.12.1.30.c.13 The emergency diesel generator preoperational test procedure was revised to require 23 consecutive valid starts, rather than 69/n (n is the number of diesels, 4), to satisfy the criteria in Regulatory Guide 1.108.
Since this i
test was not part of the Initial Startup Test Program the requirements of Facility Operating License NPF-57, License Condition 2.c(10) do not apply.
14.2.12.2 Incorporation of information previously provided in the response to Question 640.23.
14.2.12.3.5.c The control rod drive (CRD) test was revised to clarify the procedure to be used for evaluating proper CRD flow control valve response by
'dentifying station tuneup procedures rather than implying power ascension test procedures.
This change was made in accordance with Facility Operating License NPF-50, License 2.C(11) and accordingly was previously transmitted in a letter from C. A. McNeill, Jr. to T.
E. Murley dated July 21, 1986.
PAGE 21 OF 25
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SUMMARY
OF CHANGE 14.2.12.3.12.c-Reactor Core Isolation Cooling (RCIC) and High 14.2.12.3.13.c Pressure Coolant Injection (HPCI) discussions revised regarding speed and flow controller tuning for Condensate Storage Tank (CST) testing.
Discussion added regarding steam leak testing for the turbine gland seal condenser system.
These changes were made in accordance with Facility Operating License NPF-50, License Condition 2.C(11) and accordingly were previously transmitted in a letter from C. A. McNeill, Jr.
to T. E. Murley dated July 21, 1986.
14.2.12.3.12.c Changes made to clarify and distinguish between 14.2.12.3.13.c HPCI/RCIC tuneup and demonstration tests, F14.2-5, Test 13 identify approximate pump discharge pressure during CST testing, and specify the possible testing of the HPCI system in various test conditions.
These enanges were made in accordance with Facility Operating License NPF-57, License Condition 2.c(10) and accordingly were previously transmitted in a letter from C.
A. McNeill, Jr. to T. E. Murley dated November 5, 1986.
14.2.12.13.18.c Deletion of the reference to plant heat rate in the testing method f or the warranty test.
This change reflects actual contractural commitments between PSE&G and GE and did not involve a change in the Startup Test Program since the parameter was never a program requirement.
14.2.12.3.28.c Recirculation system discussion revised to accerately clarify testing of the two pump trip tests.
This change was made in accordance with the Facility Operating License NPF-50, License Condition 2.C(ll) and accordingly was previously transmitted in a letter from C. A. McNeill, Jr.
to T.
E. Murley dated July 21, 1986.
F14.2-4 The TC-4 test window has been expanded in order to provide adequate margin to scram during pressure regulator testing in natural circulation.
This change was made in accordance with Facility Operating License NPF-57, License Condition 2.C(10) and accordingly was previously transmitted in a letter from C. A.
McNeill, Jr.
to the Document Control Desk dated January 7, 1987.
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F14.2-5,; Tests 4 & 8 The full core shutdown margin and IRM performance tests were moved from Open Vessel to Heat Up to comply with the commitment that initial criticality be delayed until the reactor vessel
, head in on in accordance with the letters from C. A. McNeill, Jr. to E. Adensam dated April 8 &
10, 1986.-
This change was made-in accordance with Facility Operating License NPF-50, License Condition 2.C(ll) and accordingly was-previously transmitted in a letter from C. A. McNeill, Jr.
to.T.
E. Murley dated July 3, 1986.
F14.2-5, Test 11 Process computer testing was deleted from Heat Up since all the testing was completed during Open Vessel.
This change was made in accordance with Facility Operating License NPF-50, License Condition 2.C(ll) and accordingly was previously transmitted in a letter from C. A. McNeill, Jr.
to T. E. Murley dated July 21, 1986.
F14.2-5, Test 13 t:ote (29) was added to the HPCI test, No. 13, which discussed cold quick starts.
This change was made in accordance with Facility Operating License NPF-57, License Condition 2.c(10) and accordingly was previously transmitted in a letter from C. A.
McNeill, Jr. to T.
E.
Murley dated November 25, 1986.
F14.2-5, Tests 15, The identified System Expansion Test, Test 15, and 39 was deleted from TC-2 to reflect the actual requirements of the GE NSSS test specification as perf ormed during the Startup Test Program.
Test 39 was added to reflect the discussions in 14.2.12.3.39 and the testing performed during the Startup Program.
Both changes reflect editorial corrections to the figure and are not reflective of actual changes to the Startup Test Program.
F14.2-5, Test 26 Note 25 was added to Test 26 to permit the cold shutdown demonstration part of the test during any test condition and to achieve consistency with Section 14.2.12.3.26.c.
Secondly, an indication was added to the Test Condition 1 column to account for the hot standby demonstration part of the test.
PAGE 23 OF 25
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-SECTION
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OF CHANGE These two changes were made in accordance with Facility Operating License NPF-50, License Condition 2.C(ll) and NPF-57, License Condition 2.C(10), respectively, and accordingly were previously transmitted in letters f rom C. A.
McNeill, Jr. to T. E. Murley dated July 3 and October 2, 1986.
F14.2-5, Test 30 Move Test 30, Loss of Offsite Power, f rom TCl to TC2 and TC3 and add Note 26 to indicate testing is done with generator synchronized to the grid.
This change was made in accordance with Facility Operating License NPF-57, License Condition 2.C(10) and accordingly was previously transmitted in a letter from C. A. McNeill, Jr.
to T. E. Murley dated September 29, 1986.
F14.2-5, Test 40 Confirmatory In-plant SRV discharge testing was deleted from TC-1.
The change was made in accordance with Facility Operating License NPF-57, License Condition 2.C(10) and accordingly was previously transmitted in a letter from C.
A. McNeill, Jr. to T. E. Murley dated October 2, 1986.
F14.2-5, Test 40 Confirmatory In-plant SRV discharge testing was added to TC 3, 4, 5 and 6 and Note 27 was added indicating that testing could be performed with at least 25% power.
This change was made in accordance with Facility Operating License NPF-57, License Condition 2.C(10) and accordingly was previously transmitted by letter from C. A.
McNeill, Jr. to T. E. Murley dated October 15, 1986.
T15.0-5 Incorporation of information previously provided in the response to Question 440.33.
15.4.7.3 Incorporation of information previously provided in the response to Question 491.2.
15A.6.2 Incorporation of information previously provided in the response to Question 450.4.
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OF CHANGE Appendix 15C The.new information provided as Appendix 15C contains the Single Recirculation Loop Operation analysis which was approved with the issuance of Amendment,3 to Facility Operating License NPF-57 on August 17, 1987.
This information was previously provided in Attachment II to the letter f rom C. A. McNeill, Jr. to E. Adensam on May 30, 1986.
17.2.1.1.1 Incorporation of information previously provided F17.2-1 in the responses to Questions 260.3 and 260.13.
17.2.2 Incorporation of information previously provided in the response to Question 260.15.
17.2.6 Addition of a statement regarding the sample basis review of work procedures prior to approval which reflects the current auditing and surveillance activities performed by the Quality Assurance Department.
17.2.9 Changes made to reflect the responsibility for welding, brazing, heat tracing and NDE specifications and to utilize guidance provide in IE Information Notice 86-21.
17.2.15 Revisions made to reflect reassignment of responsibility for the review of nonconformance dispositions between the Nuclear Quality Assurance and Systems Engineering Departments.
18.1 Incorporation of information contained in SSER 1 18.2 and 2, Section 18.1 and SSER 5, Section 18.2.
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