B11111, Forwards Response to Draft SER Open Items
| ML20084E332 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/19/1984 |
| From: | Counsil W NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| TASK-1.C.1, TASK-1.C.7, TASK-1.C.8, TASK-TM B11111, NUDOCS 8405020237 | |
| Download: ML20084E332 (90) | |
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[E' j "~ HARTFORD. CONNECTICUT 06141-0270 L L y ** *v **m m acew = (203)666-6911 April 19,1984 Docket No. 50-423 Bl1111 Director of Nuclear Reactor Regulation Mr. B. J. Youngblood, Chief -
Licensing Branch No. I Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Reference:
(1) B. J. Youngblood to W. G. Counsil, Draf t Safety Evaluation Report (DSER) for Millstone Nuclear Power Station, Unit No.
3, dated December 20,1983.
Dear Mr. Youngblood:
Millstone Nuclear Power Station, Unit No. 3 NRC Procedures and Systems Review Branch Transmittal of Responses to Draf t Safety Evaluation Report Open Items Reference (1) transmitted to us the open items for Millstone Unit No. 3, including items under tne responsibility of the NRC's Procedures and Systems Review Branch (PSRB). On February 21, 1983, a meeting was neld in Bethesda, Maryland between the NRC PSRB and Northeast Utilities to discuss each of these open items. Attachment I provides the status of each PSRB open item as a result of this meeting. The status of each item is defined by one of the following ttree categories:
Closed- No f urther NNECO input or action required for resolution of item.
Confirmatory- NNECO to provide requested information on the Millstone Unit No. 3 docket at a later date. ,
Open - No resolution at this time; NNECO to address.
Attachment 11 formally transmits our respohe to each DSER PSRB open item.
The responses in Attachment 11 are being provided as they will appear in an upcoming FSAR amendment. .
1 8405020237 840419
'4/ 00 PDR ADOCK 05000423 , g ' An E PDR <j U
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t If you have any concerns related to the information contained herein or any quertions related to our responses, please contact our licensing representative, Ms. P. Capello-Bandzes at (203) 665-3714.
l Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY, ET AL By Northeast Nuclear Energy Company, Their Agent
- 4&d W. G. Counsil Senior Vice President cc: Mr. F. 3. Liederbach NRC Procedures and Systems Review Branch Mr. R. A. Becker NRC Procedures and Systems Review Branch Mr. R. Gruel Pacific Northwest Laboratory-Battelle STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD )
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Then personally appeared before me W. G. Counsll, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, an Applicant herein, that he is authorized to execute and file the foregoing l Information in the name and on behalf of the Applicants herein and that the statements contained in said information are true and correct to the best of his t knowledge and belief.
l %Ard W adilf Notary Pub My Comrmssion Expires March 31,1908 i
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o ATTACHMENTI Status of Each DSER PSRB Open item Open item Subject Status PSRB-01 TMI Action item Open*
- 1.C.1 PSRB-02 ANSI /ANS 3.2-1981 Closed (Draf t 7) or 1982, Section 5.3 l
PSRB-03 Procedures Generation Closed Package Familiarization PSRB-04 Alarm Response Procedures Closed PSRB-05 Procedures that include Closed immediate Actions to be Memorized l PSRB-06 Commitment Concerning Closed Plant Operations l PSRB-07 Procedures for Abnormal Closed Release of Radioactivity l
PSRB-08 Temporary Operating and Closed l Maintenance Procedures PSRB-09 TMI Action items Closed i I.C.7 and I.C.8 PSR B-10 ATWS Procedures Closed
! PSRB-l l Tests of Failed Confirmatory
- Fuel Monitors (Q640.2)
PSRB-12 Automatic Closure of Confirmatory
- Main Steam isolat!on Valves (Q640.3)
- NNECO considers these items to be closed, however, the NRC Reviewer has requested that we categorize these items as confirmatory, until the follow-up FSAR changes appear in a FSAR amendment.
- NNECO considers this item to be confirmatory.
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x Open Itcro 's , Subj ect ' - ' Status PCRB-13 Confornisnce to ke Confirmatory *,
Guide 1.52 (Q440.5)gulatory s s .
PSRB-14 Ceniormadce to Regulatary Corifirmatory *
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Guide 1.95 (Q640.4)
PSRB-15 Test Abstract Descriploi.s Confirmatory (Q640.7)
PSRB-16 Loss of instrument . ',
Confirmatory
- Air Test (Q640.13) ,
PSRB-17 Solid State PrStection Confirmatory.*
System (Q640,i /) -
PSR D. I Y ' ., Regulatory Guide 1.68, Open Rev. 2, Appendix A.I.2 and M.5.T (Q640.19)
PSRB-19 NReal or Dummy Fues Cobfirmatory*
Asseriuites for Vibration
' ' i cst (Q640.?O(2))
PSRB-20 NUREG-0694, item I.G.J Confirmatory *
, (Q640.22)
PSRB-21 Regulatory Guide 1.62, P.ev. 2, Confirmatbry*
Appendix A (Q640.26)
PSRB-22 Preoperational Tests Confirmatory *
'/6-84 (QG40.2?)
PSAD-23 Swing Load Test Confirmatory *
(0640.28)
Q640.15 BTP PSB-1 Confirmatory
- Q640.16 Preoperational Test Confirmatory
- Number $1 (Diehl ~
Genera sgr) .
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- NNECO considers these iten:: to be closed, however, the NRC Reviewer has requested that we categorize theie items as confirmatory, until the follow-up FSAR changes appear in a FSAR amendment.
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ATTACHMENTII Response to DSER PSRB Open items
s Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-01 TMI Action item I.C.1,"Short Term Accident and Procedures Review" (Draf t SER Section 13.5.2.2 and 13.4.2.3)
The applicant has committed to implement Supplement I to NUREG-0737 and to submit his procedures generation package (PGP) on October 1,1984, which is 3 months before the start of operator training on the Millstone Unit 3 simulator.
The PGP should be sumitted as an FSAR amendment because it provides the basis for developing the plant's EOPs. The staff's review of the PGP must be completed before issuance of the operating license and will be addressed in a supplement to the SER. Until completion of the staff review of the PGP, Task Action Plan Item I.C.1 will remain an open item.
Response
NNECO considers this to be a confirmatory item. In our April 15, 1983(l) submittal to the NRC, we agreed to provide the NRC with a procedures generation package (PGP) by October 1,1984 which will include a writer's guide, Millstone Unit No. 3 specific changes from the Westinghouse Owner's Group generic guidelines, and a description of our verification, validation and training programs. PGPs have been submitted for our Millstone Unit No. I d,3) Millstone Unit No. 214), and Haddam Neck Plant (5) which are available for NRC review.
The Millstone Unit No. 3 PGP will be similar to these in general content and format. Reviewing these documents would give the NRC a general idea of what to expect for the Millstone Unit No. 3 PGP.
(1) W. G. Counsil letter to D. G. Eisenhut, A02959, dated April 15,1983.
(2) W. G. Counsil letter to D. M. Crutchfield, A02959, dated May 13,1983.
(3) W. G. Counsil letter to D. M. Crutchfield, A03666, dated March 9,1984.
(4),(5) W. G. Counsil letter to 3. R. Miller and D. M. Crutchfield, A02959, dated September 1,1983.
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Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSRB-02 ANSI /ANS 3.2-1981 (Draf t 7), Section 5.3 (Draf t SER Section 13.5.2.2)
The applicant should modify Sections 13.5 and 13.5.1.1 of the FSAR to commit to Section 5.3 of ANSI /ANS 3.2-1981 (Draf t 7) instead of Section 5.3 of ANSI N18.7-1976/ANS 3.2, or provide justification for this deviation from the SRP. A commitment to conform to Section 5.3 of ANSI / ANS 3.2-1982 would also be acceptable, because Section 5.3 ANSI /ANS 3.2-1982 is the same as Section 5.3 of ANSI /ANS 3.2-1981 (Draf t 7). This is an open item.
Response
Refer to revised FSAR Tables 1.9-1 and 1.9-2.
The following justification is provided for the use of ANSI N18.7-1976/ANS 3.2:
- 1. Millstone Station has one common set of Administrative Control Procedures to control procedure preparation, approval, format and use.
These administrative procedures are used for all three nuclear units at Millstone. The above procedures have been developed in accordance with the Northeast Utilities Topical QA report as approved by the NRC for Millstone Units 1, 2, and 3, and Connecticut Yankee. These documents reference ANSI N18.7-1976/ANS 3.2 as endorsed by Regulatory Guide 1.33.
- 2. The NRC endorses the 1976 version of ANSI /ANS 3.2, Section 5.3 in Regulatory Guide 1.33, Revision 2 which is in accordance with Regulatory Guide 1.70, Revision 3.
- 3. Emergency Operating Procedures are developed based on the Westinghouse Owners' Group Emergency Response Guidelines as approved by the NRC.
The Emergency Operating Procedures are functional-based as described in the FSAR. Since the requirement for f unctional-based EOPs is already explicitly addressed and the approach is to comply with NRC approved procedure guidelines, little will be gained by committing to a partial standard which is not addressed within existing regulatory guides.
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() MNPS-3 ISAR I
d) TABLE 1.9-1 (Cont) 04 ,
! Y) Specific SRP Summa ry Description Corresponding SRP Section Acceptance Criteria or Dirrerence ISAR Section
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! 11.5 (Rev. 3) Table 1, item 6 - fuel storage No automatic termination 11.5.2.2.9 area ventilation system, or effluents.
Table 2, item 5 - Spent fuel No automatic termination 9.3.2
- pool treating system. , or effluents. 9.4.2 f Table 2, items 16 and 17 - Steam No automatic termination 11.5.2.3.3 generator blowdown system. or effluents.
12.2 (Rev. 2) 8.2 - Tabulation or concentrations Only normal operation 12.2.2 471.25 or airborne radioactive materials and anticipated operational 471.26 occurences are addressed
-14.2 (Rev. 2) II.4 - Categories or reportable ISAR does not provide 14.2 l occurrences that are repeatedly categories or occurrences.
I being experienced at other racilities.
- this 15.4.6 accident scenario.
i l 15.4.8 (Rev. 1) lli - Stresses should be evaluated Westinghouse considers this 15.4.8
, to escegency conditions for these a faulted condition as accidents, stated in ANSI N18.2.
l 15.6.5 (Rev. 2) 11.3 - IMI Action Plan, fio modifications have been 15.6.5.3 II.K 3.30 and 11.k.3.31. made to the small brea k LOCA model.
15.7.3 (Rev. 2) 118.1.a - Radionuclide inventory ISAR analyzed postulated tank 2.4.13.3 in railed components. failure using 1% ruel defects. 15.7.3.2 INorRT A t
Amendment 3 12 of 12 August 1983
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I FSAR Table 1.9-1 insert A 1
Specific SRP Corresponding Acceptance Summary Description FSAR i SRP Section Criteria of Difference Section j 13.5.2 C.2 - ANSI /ANS 3.2= FSAR uses ANSI N18.7 - 13.5.2 :
- 1981, Section 5.3 1976/ANS 3.2, Section 5.3 I
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. P3R B - 0 3-s MNPS-3 FSAR TABLE 1.9-2 (Cont)
- 3. SRP 11.5, Table 2, items 16 and 17 require an automatic control feature, which automatically terminates effluents of the steam generator blowdown system.
] B. Justification for differences from SRP 3
- 1. During fuel handling activities. the fuel building ventilation is processed by the fuel building filtration units. Accident
! analysis indicates that the filters prevent the release of j excessive amounts of radioactive effluent.
) 2. The spent fuel pool cooling and purification is a closed system, therefore, termination of effluents is unnecessary.
i 1 Monitoring is accomplished using the reactor plant sampling systen radiation monitor, 3SSR-RE08, and area radiation 4
monitors surveying the fuel pool. Safety evaluations described in FSAR Section 9.1.3 show this to be adequate.
- 3. !!cnitoring of the steam generator blowdown system is provided by the reactor plant sampling system radiation monitor,
- 3SSR-RE08. An evaluation of the accident scenario for a steam i
generator tube rupture shows that such an event would be
! identified by the air ejector system monitor, 3 ARC-RE21 or the main steam line monitors, 31!SS*RE75-78. The main steamline ,
. monitors would identify which steam generator is affected and l operator action would close valves to prevent release of steam
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generator blowdown effluents.
SRP 12.2 j SRP TITLE: RADIATION SOURCES
! A. Actual differences.between FSAR and SRP I SRP 12.2, Paragraph I.2 requires tabulation of the calculated
- concentrations of radioactive material, by nuclide, expected during l normal operation, anticipated operational occurrences, and accident J
conditions for equipment cubicles, corridors, and operating areas normally occupied by operating personnel. FSAR Section 12.2 does
, not tabulate the calculated concentrations of radioactive material
- expected during accident conditions.
- B. Justification for differences from SRP
, During accident conditions, local surveys and measurements will be performed as required and exposures will be limited to the requirements of NUREG-0737.
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s Amendment 3 41 of 43 August 1983 1
PS R B- O A.
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4 FSAR Table 1.9-2 Insert B I
SRP 13.5.2 1
i SRP TITLE: OPERATING AND MAINTENANCE PROCEDURES i
! A. Actual differences between FSAR and SRP The SRP references Section 5.3 of ANSI /ANS 3.2 - 1981 (Draf t 7). The FSAR is written based on Section 5.3 of ANSI N18.7 - 1976/ANS 3.2 which is endorsed in Regulatory Guide 1.33.
[ B. Justification for differences from SRP.
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- 1. The NRC endorses the 1976 version of ANSI /ANS 3.2, Section 5.3 in f Regulatory Guide 1.33, Revision 2 which is in accordance with l Regulatory Guide 1.70, Revision 3.
- 2. Emergency Operating Procedures are developed based on the Westinghouse Owner's Group Emergency Response Guidelines as approved by the NRC. The Emergency Operating Procedures are j functional-based as described in the FSAR. Since the requirement for i functional-based EOP s is already explicitly addressed, and the i approach is to comply with NRC approved procedure guidelines, little will be gained by committing to a partial standard which is not
! addressed within existing regulatory guides.
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l Millstone Nuclear Power Station, Unit No. 3 i
, Open items 1
l Procedures and Systems Review Branch i
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- PSRB-03 Procedures Generation Package (Draf t SER Section 13.5.2.2) t ,
j The applicant has stated that (a) the PGP will be submitted to NRC 3 months before start of operator training, (b) all proposed operating and maintenance i
i procedures will be completed at least 3 months before fuel loading, and (c) procedures will be available for review in advanced draf t form at least 6 months before fuel loading. It is the staff's position that procedures must be completed in sufficient time to ensure operator and appropriate plant staf f f amiliarization.
j The FSAR should describe how adequate operator and plant staff familiarization
! will be ensured. This is an open item.
! Response:
I Refer to FSAR sections 14.2.1.1 and 14.2.9. Millstone Unit No. 3 will approve and utilize its operating procedures to the extent possible to support the testing j program. This en:ures that the operating staf f is knowledgeable about the plant
- and its procedures.
- In addition, all operators undergo an extensive on-the-job systems training
- program (see FSAR Section 13.2) which requires an understanding of related system procedures.
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s Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-04 Alarm Response Procedures (Draf t SER Section 13.5.2.2) 2 The applicant should describe the system to classify or subclassify alarm responses and the methods used by operators to retrieve or refer to alarm response procedures. This is an open item.
Response
A control room annunciator response procedure is prepared to provide the operator with an immediate reference document for plant alarms. This reference document lists each control room annunciator alarm identifying its procedure reference for operator actions. Annunciator response forms are used to organize alarm lists by main board annunciator grouping, row, and column designation. Each system procedure has a designated alarm response section.
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Millstone Nuclear Power Station, Unit No. 3 l Open items !
i Procedures and Systems Review Branch t
i PSRB-05 Procedures That include immediate
- Actions to be Memorized (Draf t SER Section 13.5.2.2) 1
! The applicant should identify procedures that include immediate actions that
, must be memorized by the plant operators. This is an open item.
Response: !
l Millstone Unit No. 3 is proceeding with the development of its Emergency
, Operating Procedures based upon the Westinghouse Owner's Group Emergency j Response Guidelines, Revision 1, pending the approval of Revision 1 by the NRC. i in the event that Revision 1 is not approved by the NRC, NNECO will provide justification for the deviations from Revision 0 of the Westinghouse Guidelines.
i The procedures that include immediate actions to be memorized are:
! 1. Reactor Trip or Saf ety injection i
- 2. Loss of all A.C. Power l Immediate actions to be memorized are indicated by asterisks (*) in the j Westinghouse Guidelines.
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Millstone Nuclear Power Station, Unit No. 3
, Open items :
- Procedures and Systems Review Branch i PSRB-06 Commitment Concerning Plant Operations (Draf t SER Section 115.2.2)
Although implied in Section 115.2 of the FSAR, a clear commitment is needed
- that plant operations will be performed in accordance with written and approved ;
! procedures. This is an open item. ;
Response
- Refer to revised FSAR Section 115.2.
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P5RB - 0(o
- 3elPS-3 FSAR These procedures are reviewed and improved, if necessary to ensure ',
aparability of safety systans prior to tahng credit & the system (s) to satisfy Technical Specification requirements.
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- Special . procedures .are pewpered as necessary to support infrequent operations. The requirements for review, approval, and changes are the same as station procedures.
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13.5.2 Operating and Maintenance Procedures operating and~ malatemance procedures Mare divided ' "Into several categories which are described in the following subsections.
Table 13.5-1 lists the appropriate procedures which will initially he prepared or currently asist.
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operating and maint'enance procedures' preparation is the
- 2. - responsibility of the appropriate department head. When a procedure is written. the department supervisor will forward the procedure for review and approval in accordance with Technical Specifications. m Unit specific procedures are approved by the Unit superintendent and bg_c k common station procedures are approved by the Station Superintendent.
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-Independent position verification of safety related components / systems (valves, breakers. and control switches) with no m, indication in the control room will be performed prior to the return- .
to-service of the component / system.
l' All proposed operating and maintenance procedures will be completed at least 3 months prior to fuel loading. Procedures will be available for review in advance draft form at least 6 months prior to fuel loading. j
. 13.5.2.1 Control team Operating Procedures ;
13.5.2.1.1 General.cperating Procedures .
f These procedures cover major plant evolutions, and an initial list is included in Table 13.5-1. Step-by-step instructions are provided for the function or task with the appropriate cross reference to system aperating procedures for deta13e of specific system eperetsen.
P Appropriate precautions and limitations are included.
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. 13.5.2.1.2 system Operating Procedures
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these procedures provide step-by-step details for system operations with appropriate prereguisites, precautions, limitations, and alarm
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reopenses. Each r. f < covers the espected modes of operation of the system as well as startup, shutdown, filling and venting, and stenery operation as applicable. Table 13.5-1 includes an initial list of system aperating procedures which may be modified as I emperience dictates. .
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Insert A Plant operations will be performed in accordance with written and approved station and administrative control procedures.
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Millstone Nuclear Power Station, Unit No. 3 l
Open Items l Procedures and Systems Review Branch l
I PSRB-07 Procedtres for Abnormal Release of Radioactivity (Draf t SER Section 13.5.2.2)
A procedure or procedures coserning abnormal releases of radioactivity are not evident. This is an open item.
Response
Abnormal releases of radioactivity are addressed by administrative control procedures with respect to reporting requirements. Emergency Plan implementing Procedures (EPIP) address required actions such as radiologica!
, dose assessment (refer to FSAR section 13.3, draf t Emergency Plan, Appendix D, l Index of EPIPs). System operating procedures provide step-by-step details for system operations with appropriate prerequisities, precautions, limitations, and alarm responses as described in FSAR sections 13,5.2.1.2 and 13.5.2.2.5.
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Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch I
PSRB-03 Temporary Operating and Maintenance Procedures (Draf t SER Section 13.5.2.2)
Section 13.5.2 should be expanded to address temporary operating and maintenance procedures. This is an open item.
Response: ,
Refer to revised FSAR sections 13.5.1.3, Special Procedures and 13.5.2.2.10. ,
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- . PSRB-08 MNPS-3 FSAR
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These procedures are reviewed and improved, if necessary to ensure aperability of safety systems prior to taking credit for the #
system (s) to satisfy Technical specification requirements.
Special Procedures ELM b$YII__I b I I _. 5 _ _5 [' 5 5 .. " " " ' ' " '? N " T ' " ' " " " ""
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13.5.2 Operating and Haintenance Procedures operating and maintenance procedures are divided into several categories which are described in the following subsections.
Table 13.5-1 lists the appropriate procedures which will initially be prepared or currently exist.
Operating and maintenance procedures preparation is the responsibility of the appropriate department head. When a procedure is written, the department supervisor will forward the procedure for review and approval in accordance with Technical specifications.
Unit specific procedures are approved by the Unit superintendent and common station procedures are approved by the Station Superintendent.
Independent position verification of safety related components / systems (valves, breakers, and control switches) with no m, indication in the control room will be performed prior to the return- .'
to-service of the component / system.
All proposed operating and maintenance procedures will be completed at least 3 months prior to fuel loading. Procedures will be available for review in advance draft form at least 6 months prior to fuel loading.
13.5.2.1 Control Room operating Procedures 13.5.2.1.1 General Operating Procedures These procedures cover major plant evolutions, and an initial list is included in Table 13.5-1. Step-by-step instructions are provided for the function or task with the appropriate cross reference to system aperating procedures for details of specific system ereration.
Appropriate precautions and limitations are included.
13.5.2.1.2 System Operating Procedures These procedures provide step-by-step details for system operations with appropriate prerequisites, precautions, limitations, and alarm responses. Each procedure covers the espected modes of operation of the system as well as startup, shutdown, filling and venting, and stanty operation as applicable. Table 13.5-1 includes an initial list of system operating procedures which may be modified as I experience dictates.
13.5-4
- ' MRB~oS (FSAR pg.13.5-4)
Insert A Special procedures are prepared as necessary to support infrequently performed evolutions which will not be included in the permanent list of station procedures.
A special procedure can be written for any type of station procedure (i.e.
maintenance, operating). The form of a special procedure will be the same as the applicable type of station procedure. All requirements for review, approval, revisions, and cht.nges are the sanie as for permanent station procedures.
(FSAR pg.13.5-6)
Insert B 13.5.2.2.10 Special Procedures This topic is covered by administrative procedures. (Refer to FSAR Section 13.5.1.3, Special Procedures)
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0 NNPS-3 FSAR 13.5.2.2.3 Instrument Maintenance Instructions Instrument maintenance instructions are prepared for the performance
! of periodic calibration, testing, and channel checking of safety related plant instrumentation and all instruments used to satisfy 1
technical specification requirements. These instructions will ensure l i measurement accuracies adequate to maintain plant safety parameters within operational and safety limits. In addition, instrument !
instructions outline the periodic calibration and i maintenance accuracy requirements of test equipment necessary to support the calibration of safety related instrumentation.
13.5.2.2.4 Chemistry Procedures f
J' Chemistry procedures are prepared covering the routine analysit and ,
sampling methods to ensure compliance with plant chemistry and 4
discharge limits.
13.5.2.2.5 Radioactive Waste Procedures i
Procedures for operation of redweste systems are included in system operating procedures.
13.5.2.2.6 Plant security Instructions This topic is discussed in Section 13.6. ,
! 33.5.2.2.7 Material Control Procedures procedures j This topic is covered by administrative in i section 13.5.1.3.
! 13.5.2.2.8 Maintenance and Modificaiion Procedures i
Maintenance procedures are prepared to cover safety related work which requires a specific technique or sequence not normally part of an individual's routine skill. :
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1 The procedures support the requirements and programs of section 13.5.1.3 which covers administrative control of maintenance j and modification. i 13.5.2.2.9 Pire Protection Procedures i The Fire Protection Program is described in Section 9.5.1.
Procedures for fire protection are included under Systes Operr. ting l
Procedures in Table 13.5-1.
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Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSRB-09 TMl Action items I.C.7 (NSSS Vendor Review of Procedures) and I.C.8 (Pilot Monitoring of Selected Emergency Procedures for NTOL Applicants) in FSAR Table 1.10-1 (Draf t SER Section 13.5.2.2)
FSAR Table 1.10-1 should be revised to provide a brief explanation of how Task Action Plan Items I.C.7 and I.C.8 have been resolved, as described in Section 13.5.2.3. Suitable cross-reference between Table 1.10-1 and Section 13.5.2 of the FSAR should be provided. This is an open item.
Response
Refer to revised FSAR Table 1.10-1 and Section 13.5.2.1.4.
The Millstone Unit No. 3 Emergency Operating Procedures are being prepared based upon Revision 1 to the Westinghouse Owners' Group Emergency Response Guidelines which have not been approved by the NRC at this time. In the event that Revision 1 is not approved by the NRC, NNECO will provide justification for the deviations from Revision 0 of the Westinghouse Guidelines.
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NNPS-3 FSAR TABLE 1.10-1 (Cont)
FSAR Item and Title Position Re fe rence I.C.5 MMPS-3 meets the requirements of this item. 13.1.1 Procedures for feedback of Operating Experience I.C.6 MMPS-3 meets the requirements of this item. 13.5.1.3 Procedures for Verification or 13.5.2 Correct Performance of Operating Activities 13.3.2.1.'i 3.C.7 NSSS Vendor Review
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of Procedures 1.C.8 Pilot Monitoring of Selected -'.
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"'Kggpg 1.D.1 A control room design raview will be performed for MNPS-3 to meet the 18.0 Control Room requirements of this item.
Design Review I.D.2 A position has not been taken for HNPS-3 due to the lack of clear
- Plant Sa fe ty definition of NRC requirements for this item.
Parameter Display Console O.C.1 MNPS-3'will address this item following the issuance of NRC finalized
- T ra ining du ring Low- criteria for this item.
Power Testing j II.B.1 Safety grade reactor vessel and pressurizer venting capability is 5.4.15 i Reactor Coolant p rovided in the MMPS-3 design. 7.5 Systen vents 88.8.2 The MMPS-3 plant shielding design is outlined in Chapter 12. 3.11 Plant Shielding 12.3.2 II.B.3 MMPS-3 has a post-accident sampling system which meets the 9.3.2.6 Post-Accident requirements of this item.
Sampling II.B.4 MMPS-3 will develop and implement a training program utilizing the INPO 13.2.1 Training for guidelines for " Recognizing and Mitigating the Consequences of severe 13.2.2 Mitigating Core Core Damage" as the basis for the program.
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(FSAR Table 1.10-1) l Insert A Commitment to implement emergency operating procedwes based on NRC-approved Wes tinghouse Emergency Response Guidelines eliminates the requirements for additional NSSS vendor review of emergency operating proced tres.
(FSAR Table 1.10-1) l Insert B l
l Commitment to implement emergency operating procedures based on NRC approved Westinghouse Emergency Response Guidelines eliminates the
, requirement for pilot monitoring of selected emergency Procedtres for near-l term operating license applicants.
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( 13.5.2.1.3 Abnormal Operating Procedures Operating procedures are prepared for abnormal operation of the unit.
Abnnrmal operation is a condition that could degrade into an emergency or could violate Technical Specifications if proper action were not taken. These procedures identify the symptoms of the abnormal condition, automatic actions that may occur, and the appropriate immediate and subsequent operator actions. Table 13.5-1 includes a list of abnormal operating instructions.
13.5.2.1.4 Emergency operating Procedures Emergency operating procedures are prepared for conditions which might possibly lead to injury of plant personnel or the public if the release of radioactivity in excess of established limits occurs.
These procedures include symptoms of the emergency conditions, automatic actions that may or should occur, and immediate and subsequent operator actions. All immediste actions are required to be memorized by the operator since the primary responsibility for detection of an emergency and initiation of corrective action rests upon the operator. Emergene'/ cperating precedure: ' rill be prepar+d q ucing the eeneopt end general program-prepared-by-Westinghouse-owners V*"
sre"; "ter ::::pt:d by th; ll".C. Table 13.5-1 includes an initial list of emergency operating procedures which may be modified as experience dictates.
13.5.2.2 Station Services Procedures Station services procedures are written by the chemistry, health physics, security, quality assurance, production test, building services, training, stores, nuclear records, computer operations, station services engineering and any other station services group.
These procedures control the specific activities of these departments in support of unit or station operation (may be common site or unit specific). Statien calibration procedures uritten by the maintenance or instrument departments are also station services procedures.
Station services procedures are approved as outlined in Section 13.5.1.2. These procedures support Section 12.5 requirements. Station services procedures will be updated to reflect Hillstone 3 at least 6 months prior to fuel load. These procedures meet the requirements of Regulatory Guide 1.33, are implemented as required on two operating units, and are updated periodically.
13.5.2.2.1 Health Physics Procedures Health physics procedures support Section 12.5 and 10CFR20 requirements.
13.5.2.2.2 Emergency Preparedness Procedures Emergency preparedness procedures are covered under Section 13.3.
13.5-5
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Insert C Emergency Operating Procedures will be prepared based upon Revision i to the Westinghouse Owner's Group Emergency Response Guidelines, pending its approval by the NRC.
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- 1 Millstone Nuclear Power Station, Unit No. 3 Open Items l Procedures and Systems Review Branch PSRB-10 ATWS Procedures (Draf t SER Section 115.2.2)
As discussed in Section 15.8, " Anticipated Transients Without Scram." the ,
applicant should modif y Section 15.8 (or provide suf ficient cross-ref erence) to reflect the applicant's commitment to develop procedures for ' anticipated transients without scram based on the NRC-approved Westinghouse Emergency i Response Guidelines. This is an open item. ;
Response: 1
! Ref er to revised FSAR Section 15.8.
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15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM i
j A discussion of anticipated transients without SCRAM (ATWS) is
- ,974 is applicable to Millstone 3. t
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A 15.8.1 Reference for Section 15.8 WCAP-8330, 1974. Westinghouse Anticipated Transients Without Trip Analysis. ,
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l (FSAR pg. !5.3-1) i , insert A
! Nerthe.tst Nuclear Inergy Company (NNECO) has committed to develop procedices for anticipated transients without scram based on NRC-approved
! Westinghoust Owner's Grpup Emergency Response Guidelines (refer to section 13.5.2). .
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4 Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSitB-ll Tests of Failed Fuel Monitors - Question 640.2 (Draf t SER Section 14.2.7)
Ques tion Q640.2:
FSAR Subsection 14.2.7.7, exception l' to Regulatory Guide 1.68 (Initial Test Programs for Water-Cooled Nuclear Power Plants), Appendix A, Section Sq states that Millstone 3 does not have a f ailed f uel detection systern. FSAR Subsection 11.5.2.3.7 describes a Failed Fuel Monitor used to continuously rnonitor the reactor coolant system for f ailed f uel. Delete the exception in j FSAR Subsection 14.2.7.7 and add an appropriate test description to FSAR Subsection 14.2.12.
Response
Ref er to revised FSAR Table 14.2-1 f or a description of this test.
l This test remains as an exception because it shall be tested for proper operation prior to power ascension (instead of at power as stated in Regulatory Guide 1.68). Ref er to revised FSAR Seetion 14.2.7.7.
Additional Concerrn identified in Draf t SER:
Tests of the f ailed f uel monitor should be conducted during startup testing (at 25% and 100% power) in accordance with Regulatory Guide 1.68, Appendix A, Section 5.g.
Response
Refer to revised FSAR Table 14.2-2, Startup Test 29 f or a description of this test.
Testing of the f ailed f uel detection system will be conducted at 23 and 100 percent power in accordance with Startup Test 29.
Ref er to revised FSAR Section 14.2.7.7 f or deletion of exception to Regulatory Guide 1.68, Appendix A, Section 5.g.
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s TABLE 14.2-2'(Cont)
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- 29. STARTUP TEST - PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM
. Prerequisites for Testing '
SM Th; plcr.t i: :: :ppr:::im:t:1j 5^ p .r:;r.t pae Oe.uus 640.28 k
-Test objective and Summary This test will verify the operability of process and effluent radiation
'1.Me9.RN ' monitors.W Samples will be taken.at monitored points and analyzed.
The g results of the an.tlysis vill .be . compared to monitor.
the readings of the Acceptance Criteru The process and effluent monitor responses are consistent with sample results.
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The plant .is at approximately 50 percent power for testing of process and elfluent radiation monitors.
The plant is at approximately 25 and 100 percent power for testing of the f ailed f uel detection system.
Insert B Testing will include the f ailed f uel detection system.
Insert C The f ailed f uel detection system response is consistent with sample results.
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14.2.7.4 Regulatory C/ .
Guide 1.37, Revision 0 -
Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants 640.1 l l
The Millstone 3 initial test program will conform to the intent of Regulatory Guide 1.37.
640.1 1 14.2.7.5 Regulatory Guide 1.41, Revision 0 - Preoperational Testing of Redundant Onsite Electrical Power Systems to Verify Proper Load Group Assignments For position on Regulatory Guide 1.41, see FSAR Section 1.8.
14.2.7.6 Regulatory Guide 1.52, Revision 2 - Design, Testing, and Maintenance Criteria for Post Accident Engineered Safety Feature Atmosphere cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants For position on Regulatory Guide 1.52, see FSAR Section 1.8. -
14.2.7.7 Regulatory Guide 1.68, Revision 2 - Initial Test Programs for Water-Cooled Nuclear Power Plants The Hillstone 3 initial test program will conform to Regulatory Guide 1.68, except as specified in this section:
44 1. Th: f:il:d fuel dc tc c tin ;ys uc., (;,ppendix A, uwu y
- h:n 5: tetee s+g p =:p = t i=21 t= t. 640 2 1*-h During power escalation, testing will be conducted at the 30-percent power level instead of at the 25-percent power level. Westinghouse supplied plants have generic data for y ., .g% % the 30-percent level which they do not have at the 25-percent level (Section C.8; Appendix A, Section 5).
w c.k cAn g h Ador $--h The MSIV closure test will be performed at less than ,
44 20-percent power to demonstrate.the proper dynamic response
$1R E.S-Ah of the plant and to verify proper integrated operation of plant equipment. Plant response to a full power trip will pso,e ~'2,'h M be verified by the generator trip at 100-percent power.
ggg aw Closure of the MSIVs at 100-percent power would not provide g iT any additional information significant enough to warrant subjecting the plant to such a severe thermal transient .- ;
(Appendix A, Section 5.m.m). '
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The loss of feedwater heaters test will not be performed. Yi
- I Since plant response to load swings and large load I reductions is demonstrated in other tests, there is no need I to subject the plant to this additional transient (Appendix A, Section 5.k.k).
{ 5, Millstone 3 does not have a partial scram feature (Appendix A, Section 5.j).
1 Amendment 5 14.2-17 November 1983 1
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Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSRB-12 Automatic Closure of Main Steam Isolation Valves -
Question 640.3 (Draf t SER Section 14.2.7)
Question Q640.3:
Your exception to testing the automatic closure of all main steam isolation
- valves (FSAR Subsection 14.2.7.7(3)) at 100 percent power does not supply adequate technical justification for conducting the test at a low power level.
Provide adequate technical justification or revise the FSAR to indicate that the test will be conducted at f ull power.
Response
FSAR Sections 15.2.3.1 and 15.2.4 indicate that the dynamic response of the plant to a MSIV closure is bomded by the response of the plant to the turbine trip event because closure time for turbine stop valves is faster than MSIV's.
FSAR Section 15.2.3.2 describes the LOFTRAN code used to model the turbine trip event. This program does not take credit for steam dump. Plant response to a turbine trip from 100% power will be demonstrated per Startup Test 39 (FSAR Table 14.2-2). The combination of an MSIV trip at 20% power, the LOFTRAN model, and a 100% power turbine trip (with normal steam dump operation) should provide adequate verification of plant response to this transient.
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Millstone Nuclear Power Station, Unit No. 3 Open Items Procedtres and Systems Review Branch PSRB-13 Conformance to Regulatory Guide 1.52, Paragraphs C.2.1, C.11, and C.1p. - Question 640.5 (Draf t SER Section 14.2.7)
Question Q640.5:
Certain exceptions to Regulatory Guide 1.140 (Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants) listed in FSAR Section 1.8 need to be deleted or modified as described below to be acceptable.
- 1. Modif y exception to Paragraph C.2.f to delineate how the ductwork leak tests performed using the methods of the Associated Air Balance Comcil differ from the requirements given in Section 6 of ANSI N510:1975, and provide technical justification for any testing that does not address those differences.
- 2. Modify exception 1 to Paragraph C.11 to provide assurance that the data provided in the certified f an performance curves will most closely represent the manner in which the f an will be installed in the appropriate system.
1 Modify exception 2 to Paragraph C.li to either ref erence the displacement criteria that will be used, or agree to meet the criteria given in the 1980 revision to ANSI N509 Section 5.7.1
- 4. The exception to Paragraph C.11 states that an exception is taken to the following: " Class B leakage rates shall be determined for one damper of each type instead of every damper." If the intent is to not test each damper's leak rate, expanded technical justification will be required and the exception rewritten to clarif y what is actually intended.
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- 5. Modify FSAR Subsection 14.2.7.15 to either include the exceptions listed in FSAR Section 1.8 or to ref erence FSAR Section 1.8.
Response
Ref er to revised FSAR Section 14.2.7.15 and revised Table 1.8-1 ior the response to this question.
Additional Concerns Identified in Draf t SER:
The exceptions noted in this item should also be ad&essed in the statement of conformance to RG 1.52, Paragraphs C.2.1, C.11, and C.1p.
Response
Refer to revised FSAR Table 1.8-1.
e n.f4PS-3 iSAR 1 Allt i 1.8-l ( Corit )
b R.C.
Deinree o' Comp.Isance ISAH Section he t e rence f4A istip f e l ter components ori a cell by cell basis.
@ hemisters, heaters, fans and casangs M wall be decontamsnaled by wash down O proress; wash slown Isquid will d ra i re
'A to are aes a ted d ra in system.
Paray nsph C 4 5
!!ousing leak tests are performed a n accor-dance willi the provs sions saf Sect ion 6 of' ANSI P4'a ln- 19,"; a s r ecommensted in this paraysaph, lloweve r, duc two rk Iests a v et per f ormed tosing acceptable metliods of the Associa ted Ai r lla lance Counci l, ec he .- -
Q Panty r.iph C. 3gi_{ cia ri ficalion)
All tilPA filters ase shapped to an NHC (Juality Assurance Station for testing.
Iloweve r , if data confirm that HIPA filters are damaged by the additional trans-portateon, and/or time handlirig at the NRC f'a c i l i t y, the decisiore to send all HfPA f s I ters f or add a t aonal testing wiII be r ecosis o de red . If'IILPA filters are not sent to the NitC 4taa li ty Assurance Stat ion, 3
- ,uf fic ient addi tiona l testing remaires to enster e lif PA f s I ter rol sahi i e ty. Ihe lif PA fil ter cell testing is conducted iriitially at the manufacturer's f aci li t ies and again af ter installation at the plant site. All lif PA l'il ters furni shed a re equipped wi th f ace qua rds in accordance with HIL-F-51068. When installed in the fi l ter housing, the lifPA filters and liousing are irispected for defects and tested for leak Lightness in accordance with ANSI N'210- 19 75.
Pa ragraph C. 3.e (Cla ri fication]
f il ter and adsorber mounting f rames a re constructed and designed in accordance with the recommendations of Section 1.3 4 o f E 4DA 16-21 except for the frame tolerance guidelities in Table 8 6.P. The tolerances selected for HLPA and adsorber mountings are sufficient to satisfy the bank leak test criteraa of Pa ragraphs C.S.c and C.$.d of Hegu-18 of 58 August 1983 Amendment 3 b.
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MNPS-3 ISAR TABLE 1.8-1 (Cont) p)
e ISAR Section 00 R.C. Reference Decree of Compliance ng Hot Title C latory Guide 1.52 Rev. 2.
P-fa rag raph C. 3.q Hillstone 3 is in accordance with ANSI N509, except access to the control building filter units is not provided with hinged doors or inspection windows. Access is via 20-inch by 40-inch bolted panels. Other units are provided with hinged doors or bolted panels with inspection windows. There is no internal lighting.
Pa raa raph C. 3.h Exception is taken to the recommendations of Section 4.5.8 of [RDA 76-21 relative to drain sizes and arrangement. Normally closed manual valves, instead of water seals and traps, will be provided to control the discharge of the fire sprinkler flow. Sprinkler flow will be a timed discharge, and the water will be con-tained within the housing until it is re-moved to the liquid radwsste system at a controlled rate.
Pa raqqaph C. 3.1 The dwell time for the minimum 2 inches of the carbon adsorber unit is 0.25 sec. for bed depths greater than 2 inches, where the dwell time is less than 0.2 sec per 2 inches of total bed depth, experimental verification of fil-ter assembly will be provided.
Pa raa raoh C. 3. k When conservative calculations show that the maximum decay heat generation from collected radioiodines is insuf ficient to raise the ca rbon bed temperature above 250*f with no system overflow, small capacity EST atmosphere cleanup systems may be designed without an air bleed cooling mechanism.
Exception is taken to the requirement of any cooling mechanism satisfying single-failure criteria because a backup mechanism is pro-vided.
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NNPS-3 FSAR TABLE 1.8-1 (Cont)
R.C. FSAR Section No. Title Decree of Compliance Re fe rence in addition, exception is taken to provide
- 0) hundity control for the decay heat removal Of system cooling air flow which uses room air 43 of less than 70 percent relative humidity.
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Pa rag raoh C. 3.1 System resistances will be determined in accordance with Section 5.7.1 of ANSI N509-1976 except that fan inlet and outlet losses will not be calculated in accordance with AMCA 201. Fan blast area data necessa ry to calculate inlet and outlet losses, per AMCA 201, are the responsibility of fan manufac-turers, and are not available from them.
Erc ;t! n ! i: Err 1: E :t! n 5.'.2 cf A%00
%500 : Of cf c0ar:
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. 79, e n r. nn, ting; cr tGst-ry uhon ca re i rlad_fa n
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Ex::;t!:n !: ::Ern : 5:!:nc!n; teeha+ ques d:r'n d '- E::: :n 5.7.3 Of ANE: %500-1976. I r
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indu;;ry pr :;l:: U. L: u;;d sh n ;;x!;u;
.!tr:t!:n ;;:: !ty 20th:f Ir;:::: 2nrc::!: tic re; ! : Trt :t certe! ;;;r: tin p :d .
Documentation will not be furnished in accordance with Section 5.7.5 Where AMCA certification ratings are submitted.
Pa rac ra ph C. 3. n Exception is taken to Section 5.10.3.5 of ANSI N509-1976: ductwork, as a structure, will have a resonant frequency above 25 Hz, but this may not be true for the unsupported plate or sheet sections. The design provides for specification of the resonant frequency range of the support hangers. Specifying the resonant f requency of the unsupported plate or sheet has no meaning in the design.
Pa rac raoh C. 3. 0 Exception is taken to the provisions in Section 5.9 of ANSI N509-1976 of designing dampers to ANSI B31.1 and to using butter-20 of 58 C
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R.C. ISAR Section
!!ou lille Degree of Compliance Re fe reace fly valves. Class H dampers may tut designed atul tested to meet the verification of stsength ar.d leaktightness necessa ry for use in a containment air stream. (Note: This exception does not pertain to containment penetrations.)
in addition. exception is taken to the followsng:
Class B leakage rates shall be deter-mined for one damper of each type instead of every damper.
Pa rag raph C.4.a faception is taken to full compliance with Section 2.3.8 of ERDA 16-21 i.e., the plant does not use any communications system, floor drains are as noted in Pa ragraph C. 3.h above, decontamination areas and showers are not "nea rby," fi l ters a re not used at duct inlets, and duct inspection hatches are not provided.
Pa raoraph C.4.b Partial compliance, with a minimum spacing be tween f i l te r f rame o f 2 f t-6 in. instead of a minimum of 3 feet. This is deemed adequate since 3 replacement of filter elements would be minimal due to system function, use, and location.
Pa raoraph C.4.d (Cla rificat ion)
IST atmosphere cleanup systems are run a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month, lioweve r, if the field data confirms that it is unnecessary to run the trains 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month to reduce the amount of moisture present on the fil-te rs, this decision will be reconsidered.
1.533 Application of the Single- Comply, wi th the following cla ri fications: 3.1.1 f ailure Criteriors to Nuclear Power Plant Protection Systems 1 Regulatory Position C.1 (Rev. O, June 1913)
Due to the trial-use status of the
. source ducument. I f f E 3 79-19 77, depa r-
,, ture f rom certain provisions may occur.
- The phrase "any and all combinations of Ahendment 21 8 August 19
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FSAR Table 1.8-1
, insert A Ductwork leak testing is performed using the direct measurement method.
i Meastrement apparatus included a blower, calibrated orifice, and manometer.
Since the only ductwork utilized on ESF air cleaning systems is classified as Leakage Class II, this method provides equivalent acctracy and ANSI N510 methods.
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Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-14 Conf ormance to Regulatory Guide 1.95, Position C.5 - Question 640.4 (Draf t SER Section 14.2.7)
Question Q640.4:
If you intend to conduct an initial control room gross leakage rate test as part of the preoperational test program, delete FSAR Paragraph 14.2.7.11. FSAR Section 14.2.7 should be limited to discussion of Regulatory Guide exceptions relating to the initial test program.
Response
Ref er to revised FSAR Section 14.2.7.11 for the response to this question.
Additional Concerns identified in Draf t SER:
The statement of conformance to RG 1.95, Position C.5, should be included in FSAR Subsection 14.2.7 and should only discuss those exceptions relating to the initial test program,if any.
Response
Ref er to revised FSAR Section 14.2.7.12.
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,.3 conditions will be verified by analysis based on as-built HPSI pump
( and system head-capacity curves; however, the operability of the check valves will be demonstrated by testing. Power system response to a safety injection signal will be verified during other testing (Section m C.1.a.(2)).
640.4 14.2.3 [ Regulatory Guide 1.106, Revicion 1 - Periodic Testing of 13 Diesel Gennator Units Used as Onsite Electric Power Systems at 11uclear Power Plants For position en Regulatory Guide 1.108, see FSAR Section 1.S.
Regulatory Guide 1.116, Revision 0-R - Quality Assurance 640.4 l14.2.7 g Requirements for Installation, Inspection. and Testing of llechanical Equipment and Systems The Millstone 3 initial test program will conform to the intent of Regulatory Guide 1.116.
640.4 l14.2.7hRegulatory Guide 1.12S. Revision 1 - Installation Design g and Installation of Large Lead Storage Batteries for Nuclear Power Plants The Hillstone 3 initial test program will conform to the intent of Regulatory Guide 1.128.
640.4 l14.2.7[/ Regulatory Guide 1.140 Revision 1 - Design, Testing, and <
g !!aintenance Criteria for !1ortral Ventilation Exhaust System Air Filtration and Absorption Units for Light-Water-Cooled Nuclear Power Plants 640.5 lForpositiononRegulatoryGuide1.140,seeFSARSection1.8.
14.2.8 Utilization of Reactor Operating and Testing Experience in Development of Test Program The Hillstone 3 test program will utilize information gaine3 from operating and testing experience at similar nuclear plants to p. ovide guidance in developing test procedures and schedules and to alert personnel to potential problem areas.
The !!illstone 3 Superintendent will designate individuals on the plant staff to revieu pertinent industry literature, such as NRC IE bulletins, circulars and information letters, vendor information notices and applicable event reports from other facilities.
Commitments resulting from this review will be tracked to ensure incorporation into plant procedures or design.
Amendment 5 14.2-20 November 1983
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________. J
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(FSAR pg.14.2-20)
INSERT 14.2.7.12 . Regulatory Guide 1.95, Revision 1 - Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release.
For position on Regulatory Guide 1.95, see FSAR Section 1.8.
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Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch l
l PSRB-15 Test Abstract Descriptions -
l Question 640.7 (Draf t SER Section 14.2.12)
Question Q640.7:
Regulatory Guide 1.70 Paragraph 14.2.12 states that test descriptions should include a " summary of acceptance criteria." To comply, you should include, for all tests listed below, acceptance criteria or a discussion of the sources for the acceptance criteria to be used when test procedures are prepared. This information is necessary for the NRC inspectors who review test procedures and evaluate test results. The test description should provide " traceability" to acceptance criteria sources such as: specific FSAR Subsections, Technical Specifications, topical reports, vendor-furnished test specifications, and/or accident analysis assumptions.
- 1. Preoperational Test Numbers 1-11, 13-14, 16-29, 31-60, 62-68, 71, and 73-75.
- 2. Startup Test Numbers 1-2, 7-8,11- 13,17-19, 22-24, 26, 28-35, and 38.
Response
The sources for acceptance criteria will be provided in the individual test procedures when they are prepared. Traceability to specific FSAR subsections, Technical Specifications, topical reports, vendor-furnished test specifications, and/or accident analysis assumptions will be provided.
Approved test procedures for satisfying FSAR test commitments will be made available to the NRC staff personnel from the office of Inspection and Enforcement approximately 60 days prior to their intended use as required by Regulatory Guide 1.68 or approximately 60 days pricr to fuel load, whichever is sooner.
Additional Concerns identified in Draf t SER:
Test abstract descriptions should be expanded to indicate the sources of acceptance criteria.
Response
Refer to FSAR Table 14.2-3.
F5RB- 15 TABLE 14.2-3 PREOPERATIONAL/ ACCEPTANCE /STARTUP TESTS ACCEPTANCE CRITERIA SOURCES Test Number Title Sources P1 RCS Cold Hydro FSAR Table 5.4-15 P2 Control Rod Drive FSAR 4.6.3; Vendor Specification 001 (Westinghouse)
P3 Fuel Transf er Vendor Specification 001 (Westinghouse)
P4 Polar Crane Vendor Specification 014 (Harnischfeger)
P5 Volume Control Westinghouse (W) NSSS SU Manual (Charging and (NEU-SU-2.C3); W Precautions, Limitations, and Letdown)
Setp(oints (PLS)Engineering) 459 Combustion Vendor Specifications 001 (W) and P6 Volume Control (Boric Acid) NEU-SU-2.2.3; IEB 81-02 P7 Volume Control (BTRS) NEU-SU-2.2.3; W PLS P8 Fuel Pool Cooling FSAR 9.1.3 P9 Containment Recirculation FSAR 6.2.2.3 P10 Residual Heat Removal FSAR 5.4.7, 6.3 Pil LP Saf ety injection FSAR 6.3; R.G.1.79,1.108 P12 HP Safety injection FSAR 6.3; R.G.1.79,1.108 P13 Quench Spray FSAR 6.2.2; R.G.1.1,1.26,1.29,1.97 P14 Reactor Plant Sampling FSAR 9.3.2, 9.3.4 P15 Containment Local Leak FSAR 6.2.4, 6.2.6; Table 6.2-4; Rate Testing 10CFR50 Appendix 3 P16 Containment Ventilation FS AR 6.2.5.4, 9.4.7, 9.5.10.4 Pl7 Auxiliary Bldg. Ventilation FSAR 9.4.3.1 P18 Waste Building Vent FSAR 9.4.2, 9.4.9.1 P19 Fuel Building HVAC FSAR 9.4.2, 9.4.9.1 P20 ESF Building HVAC FSAR 9.4.5 P21 Control Building HVAC FSAR 6.4.3, 6.4.5, 9.4.11 R.G.1.95
. _ . - . _ .=_- _- _ _ - . - - _- . -. _ - . . . . . . . - - _ - _ . . - --.
p3R6-Iy l =
1 Test l Number Title Sources j i ,
l P22 Screen House HVAC FSAR 9.4.8.1.1 i i P23 EGE Vent FS AR 9.4.6.1.3, 9.4.6.5 l r P24 Supplementary Leak FSAR 6.2.3.3 i i Detection and Release l
j P25 Main Steam FSAR 10.3.3; NEU-SU-2.8.3, 2.8.5
.! P26 Steam Dump Control N EU-S U-2.8.3, 2.8.5 l P27 Steam Generator Blowdown FSAR 10.4.8 i l
I P28 Main Feedwater FSAR 10.4.7; Vendor Specification f 021 (General Electric) ;
P29 Steam Generator % ater Level FSAR 10.4.7.2 [
] Control ;
! P30 Auxiliary Feedwater FSAR 10.4.9 l
! P31 Service Water FSAR Table 9.2-1
- P32 Reactor Plant Component FSAR Table 9.2-5 Cooling P33 Reactor Plant Chilled Water FSAR 9.2.2.2.1, Table 9.2-7 '
1
] P34 Charging Pump Cooling FSAR 9.2.2.4.2, Table 9.2-10 P35 51 Pump Cooling FSAR 9.2.2.5.2, Table 9.2-12 l P36 NST Cooling FSAR 9.2.2.3.2 !
i i P37 Reactor Plant Gaseous Drains FSAR 9.3.3; R.G.1.70 t i
l P38 Instrument Air and FSAR 9.3.1.1.4.1; R.G.1.68.3 3
- Containment Instrument Air l
P39 Rad. Liquid Waste FSAR 9.3.3,11.2,11.5; R.G.1.70 l P40 Baron Recovery FSAR 9.3.5.1
! P41 Rad. Gaseous Waste FSAR 11.3 I P42 Rad. Solid Waste FSAR 11.4 P43 Steam Generator Chemical Feed FSAR 10.4.7; Vendor Specification 053 4
(Yarway)
P44 Fire Protection - Water FSAR 9.5.1 i
.i e
i P3Rer-IG Test Number Title Sources P45 Fire Protection - CO2 and FSAR 9.5.1 HALON l P46 4KV Normal and Emergency FSAR 8.3.1.1, Table 3.1-2 Distribution A/P47 480V Normal and Emergency FSAR 8.3.1.1 Distribution i P48 120 VAC Instrumentation Vendor Specification E261 (Solidstate Controls) l Non-Vital Distribution i
l P49 120 VAC Instrumentation Vendor Specification E622 (Elgar)
Vital Distribution P50 125 VDC Distribution FSAR 8.3.2.1, Table 8.3-5; Vendor Specification E262 (GE)
P51 Diesel Generator FSAR 8.1.7, 9.5.6.1; R.G.1.79,1.108 P52 Diesel Generator Fuel FSAR 9.5.4 l P53 RSST FSAR 8.3.1.1, Table 8.1-2 P54 Communications FSAR 9.5.2; IEB 79-18 P55 Nuclear Instruments Westinghouse PLS P56 Incore Nuclear Vendor Specification 001 (W)
Instrumentation P57 Process and Area Rad. FSAR Tables 11.5-1,2; 12.3.4 Monitoring P58 ESF Actuation (Diesel Sequencer) FSAR 8.3 P59 Reactor Trip (Solid State FSAR Table 15.0-4; W PLS Protection) l P60 Process Protection and Control Vendor Specification 001 (W)
~
Instrumentation Racks l
P61 Protection /Saf eguards System FSAR Chapter 15 Response Time P62 DRPI Vendor Specification 001 (3')
l P63 Loose Parts Monitor FSAR 4.4.6.4 l
P64 Seismic Mcnitor Vendor Specification 319 (Terra Technology)
I
l
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POR B - 16 f Test Number Title Sources l
P65 Emergency Lighting FSAR 9.5.3 P66 ESF Integrated Test w/o FSAR 7.3; R.G.1.79 Loss of Normal Power P67 ESF Test with Loss of FSAR 8.3.1.1.2.4; R.G.1.108 Normal Power P68 Leak Detection Technical Specifications; R.G. 1.79,1.10 P69 Containment Isolation FSAR 6.2.4 P70 Containment Integrated FSAR 6.2.6; ANSI N45.4; 10CFR50 Appendix 3 Leak Rate P71 Integrated Precore Hot R.G .1,68,1.79 Functional Testing P72 Reactor Coolant and Associated
- Systems Expansion and Restraint P73 Reactor Coolant and Selected NETM-50 Systems Piping Vibration P74 Thermal Expansion of Piping NETM-50 and Components of Secondary Systems P75 Control System Test for N EU-SU-2.74, 3.13 Turbine Runback P76 RCIV Vendor Specification 001 (W)
P77 Condensate and Condensate
- Storage A78 Turbine Plant Sampling FSAR 10.4.7.4 A79 Turbine Plant Component
- Cooling A/P30 Heat Tracing IEN 79-24 P81 RWST Cooling
- P82 Reactor Vessel Head Vent
- and Transfer Tube
,_ - .-- = _ - - - - - - - . - _ . -. .- __ -- . -.
PSRB- IF i e l l
Test Number Title Sources S1 Initial Core Load W Nuclear Design Report S2 Post-Core Hot Functional See individual tests S3 CRDM NEU-SU-2.5.1; R.G.1.68 54 RPI NEU-SU-2.5.4 S5 Rod Drop Times NEU-SU-2.5.3; Technical Specifications; R.G.1.68 56 Rod Control System NEU-SU-2.5.2 57 Pressurizer Spray and Heater NEU-SU-2.1.5 Capacity 58 RTD Bypass Loop Flow N EU-SU-2.1.9 59 Reactor Coolant System Flow FSAR Table 4.4-1, Technical Specifications S10 Reactor Coolant Flow Coastdown FSAR 10.3 511 Movable Incore Detectors N EU-SU-2.9.3 S12 Operational Alignment of NEU-SU-2.9.6; R.G.1.68 Process Temperature Inst.
513 Computer Programs Baseline data aquisition 514 Vibration and Loose Parts
- Monitoring S15 Water Chemistry Control
- S16 Radiation Survey FSAR 12.3.1 517 Initial Criticality Technical Specifications S!8 Low Power Physics Test W Nuclear Design Report S19 Boron Reactivity Worth W Nuclear Design Report 520 Pseudo Rod Ejection FSAR 15.4 S21 Natural Circulation FSAR 14.2.12.2,15.2.6; R.G.1.68 522 Power Ascension R.G.1.68 523 Dynamic Automatic Steam NEU-SU-2.8.5 Dump Control
~ -. _ . -_ --. - _. . .
I PSR&l5 Test Number Title Sources
- 524 Auto Steam Generator Level NEU-SU-2.8.2
- Control S25 Shutdown from Outside Control R.G .1.68.2 Room j S26 Station Blackout R.G.1.68 r
S27 MSIV Closure FSAR 10.3.3 i
S28 Operational Alignment of W PLS Nuclear Instrumentation S29 Process and Effluent FSAR 12.3.4, Table 11.5-1,2 Monitoring 530 Core Performance Technical Specifications; R.G.1.68 l S31 Power Coef ficient NEU-SU-2.9.11 S32 Axial Flux Dif ference Technical Specifications Instrumentation Calibration S33 Ventilation Systems Operability FSAR 9.4, Table 9.4-1; R.G.1.68 ;
S34 Turbine Generator and Baseline data acquisition Feedwater Turbine Operability 535 Calibration of Steam and N EU-SU-2.9.4 Feedwater Flow Inst.
I j S36 Auto Reactor Control NEU-SU-2.8.1; W PLS i
j S37 Load Swing NEU-SU-3.4.7, 3.4.8 ,
) S38 Auxiliary Coolant Systems FSAR 9.2.2, 9.2.7 Performance
, 539 Unit Trip From 100% Power FSAR 15.2.3; R.G.1.68 ,
S40 Warranty Run NEU-SU-3.5.1 i
541 Secondary Plant Performance
- i i S42 Containment Penetration
- Temperature Monitoring Note: This listing is only a partial summary of the acceptance criteria sources used to prepare the indicated test procedures. A detailed listing will be available in i cach test.
- The sources of acceptance criteria for these tests can be found in the test abstract descriptions.
4
. s Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-16 Loss of Instrument Air Test -
Question 640.13(Draf t SER Section 14.2.12)
Question Q640.13:
FSAR Subsection 9. 3.1.1.4.1 states that while the instrunent air system is not safety related,it does have an interface with components that are part of saf ety related systems. Modify Preoperational Test Number 38 (Instrument Air and Containment Instrument Air), or the preoperational test objectives in FSAR Table 14.2-1 for all safety related systems that interf ace with instrument air, to include individual valve testing in accordance with Section C.8 of Regulatory Guide 1.68.3 (Preoperational Testing of Instrument and Control Air Systems), or revise the carent exception to Regulatory Guide 1.68.3 in FSAR Subsection 14.2.7.9 to include a listing of the applicable saf ety related systems.
Response
Refer to revised FSAR Table 14.2-1 for the response to this question.
Additional Concerns Identified in Draf t SER:
The loss of instranent air test should be conducted to simulate both a gradual loss of presstre as well as a sudden loss of presstre.
Response
Ref er to revised FSAR Table 14.2-1, Preoperational Test 38.
Air-operated valves will be tested to verif y f ailure position on a sudden loss of air pressure on an individual basis. In addition, dtring hot imctional testing, a gradualloss of instrument air test will be performed as Freoperational Test 38 in accordance with FSAR Section 9.3.1.1.4.1 and Regulatory Guide 1.68.3.
. P9RB - l(o
, ., NNPS-3 FSAR TABLE 14.2-1 (Cont)
- 38. PREOPERATIONAL TEST - INSTRUIIENT AIR AND CONTAIN!!ENT INSTRUI Prerequisites for Testing General prerequisites have been met. The system has been pressure tested using instrument air quality gas.
Test Objective and Summary Testing will be performed to provide assurance that the instrument air system will provide clean dry air at the proper pressure to end use equipment.
All air operated valves are individually tested to ensure proper operation. This testing includes proper response to loss of air. 640.13 Compressors will be tested for manual and automatic starting, quality and volume of air delivered and verification of instrument readings.
Cooling water flows to the compressors will be verified. Instrument air dryers will be coupled to the compressor and full flow air tests will be conducted.
of dryer towersDryers will be operated full cycle with automatic switching verified.
verified. Instruments and alarm settings will be leakage from Total theair demand system, willat benormal steady state conditions, including design. verified to be in accordance with Quality of air will be evaluated at the dryer outlet. Further verification of cleanlinese shall be verified by blowdown of instrument C gg air lines through a filter cloth. A41oss of instrument air test shall be conducted at near normal operating conditions to verify acceptability of emergency response procedures and system response. A test shall be conducted to demonstrate that plant equipment designed to be supplied by the instrument less restrictive air quality requirements. air system is not supplied by other air sup large quantities of instrument air Plant components requiring while the system is at near normal steadyshall be operated state simultaneously conditions to verify i that pressure transients in the distribution system do not exceed acceptable values.
Functional testing shall be parformed to verify that failures resulting in an increase in the supply system pressure will not cause peak transient pressures above the design pressure of the system components.
Acceptance Criteria All equipment in the instrument air system will perform in an acceptable manner in accordance with design requirements.
Amendment 4 41 of 82 September 1983 1 l9, u s 's,',,s *3
,4
Millstone Nuclear Power Station, Unit No. 3 i Open items j Procedures and Systems Review Branch l
i
! PSRB-17 Solid State Protection System -
Question 640.17 (Draf t SER Section 14.2.12) ;
i
- Question Q640.17
Modify Preoperational Test Number 59 (Solid State Protection System) to provide asstrance that a manual reactor trip will both remove voltage from the under-voltage trip coil and energize the shtsit trip coli (see !&E Bulletin 83-01, ;
February 25,1983).
Response
Refer to revised FSAR Table 14.2-1, Preoperational Test 59. l 1
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MNPS-3 FSAR
( TABLE 14.2-1 (Cont)
- 59. PREOPERATIONAL TEST - SOLID STATE PROTECTION SYSTEM Prerequisites for Testing -
General prerequisites have been met.
Test Objective and Summary Testing will demonstrate proper operation of the reactor trip and i
engineered safeguards actuation logic and output signals of the solid
.) state protection system in response to simulated input signals on each l channel. Each design logic condition will be tested and proper j coincidence logic verified. Fail safe operation on loss of power will 3
be verified. The manual reactor trip up to the tripping of the reactor trip breakers will also be tested.
l N AT Acceptance Criteria The solid state protection system produces proper logic response for specified input signals.
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63 of 82
. PORB-l9 i FSAR Table 14.2 l Preoperational Test #59 Y
INSERT 1
r i I j This will include testing to individually test that a manual trip will remove power ;
1 4
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Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-18 Regulatory Guide 1.68, Rev. 2, Appendix A.I.2 and A.5.t - Question 640.19 (Draf t SER Section 14.2.12)
Question Q640.19:
Regulatory Guide 1.68, Revision 2, Appendix A.I.2 and A.5.t prescribe testing for various valves. Modify Preoperational Test Number 71 (Integrated Precore Hot Functional Testing) to provide for a more complete demonstration of the operability of pressurizer power operated relief valves; main steam line relief valves; atmospheric steam dump valves; main steam bypass valves; and main steam control valves. Such a demonstration should include response times, relieving capacities, setpoints, and reset pressures. Open and reclosure setpoints for all relief valves should be checked at temperature. Where valves are not tested in-situ with the process fluid, testing should be conducted to verify that i
discharge piping is clear and will not choke or produce back-pressure affecting set-reset pressures of the valves. When referencing bench tests instead of performing installed capacity checks, technical justification should be provided.
(NOTE: This item is not applicable to ASME Code safety valves subject to ASME Section XI preservice tests.)
Response
Refer to revised FSAR Table 14.2-1, Preoperational Test 71.
These valves will be tested at temperature, in place, with the process fluid. The relief capacity of the atmospheric dump values and mnin steam bypass valves is demonstrated in tests described in FSAR Table 14.2-2 test numbers 23, 37, and
- 39. The capacity of the PORV's is addressed in FSAR Section 5.4.13.2. ASME Code Safety Valves, will be subject to ASME Section XI testing.
Additional Concerns identified at March 21,1984 Meeting:
Relief capacity of the PORVs and atmospheric dump valves has not been adequately demonstrated. FSAR Subsection 5.4.13.2 addresses an evaluation program whose results will be reported to the NRC prior to f uel load, and startup tests 23, 37, and 39 do not specifically address determination of valve relief capacity. Provide reference to where specific testing is accomplished which ensures that the relief capacity of the PORVs and atmospheric dump valves is less than the value assumed in the safety analysis (FSAR Subsections 15.1.4 and 15.6.1).
Response
Refer to revised FSAR Table 14.2-1, Preoperational Test 71.
. _ _ _ _ _ . . ~ _ . _ . _ . _ ___.._4 - . _ _ _ . _ _ . _ . _ . _ ._.__
0 /
! Millstone' Nuclear Power Station, Unit No. 3
) '
Open Items
} ,
Procedures and Systems Review Branch 5
PSRB-13 (Cont )
- Although specific testing is not performed to ensure that PORV and atmospheric l dump valve relief capacity is less than the value assumed in the safety analysis, the following system design limits npply which limit the effect of excessive i relief capacity: t l
l approximately 30% of what a pressurizer safety valve would relieve. l Bench test result <> . provided by EPRI indicate that PORY relief f
- capacity is 372,600 lbm/hr, while safety valve design relief capacity l j is 420,006 lbm/hr (refer to EPRI NP-2628-LD, dated September, .
- 1982). Since the RCS is analyzed for inadvertent safety valve [
j opening, inadvertent PORV opening is, therefore, bounded by the j
' former event. ,
- 2. Inadvertent atmospheric dump valve opening is bounded by the l l analyzed main steam line rupture event due to relative pipe sizing. l FSAR Section 15.1.5.2 states that the Main Steam Line Rupture i
! event is analyzed for an equivalent 1.4 f t2 break. Since failure of an !
- Atmospheric Steam Dump would result in a maximum of a 0.35 f t 2 l l
break, this accident is bounded, i _
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P5RS - I S MHPS-3 FSAR g TABLE 14.2-1 (Cont)
V 71. PREOPERATIONAL TEST - INTEGRATED PRECORE HOT FUNCTIONAL TESTING Prerequisites for Testing General prerequisites have been met. The reactor coolant system cold hydrostatic test has been completed. All preoperational testing of systems required to, support hot plant operations has been completed and reviewed for adequacy for the joint test groups with all test deficiencies corrected or specifically waived. ,
Test Objective and Summary Testing will demonstrate the satisfactory performance of systems and components during the heatup of the reactor coolant system (RCS),
operation at normal temperature, pressure, and cooldown. Specific testing will include:
- 1. Heatup of the RCS to normal operating temperature and pressure utilizing the reactor coolant pumps and pressurizer heaters.
This test will include demonstration of solid system pressure control and the capability to add hydrazine to the RCS
- 2. Perform periodic vibration measurements of reactor coolant pumps O' 3. Demonstrate that the operation of pressuricer pressure and level control systems including heater and spray operation.
Perform preliminary spray flow adjustments
- 4. Demonstrate that the operation of the steam generator atmospheric and condenser steam dump valves is acceptable g g g 7" within specific limits 4 - ,
Demonstrate the capability of the chemical and volume control 5.
system to provide charging water at rated flow against normal RCS pressure, verify letdown flow rate for various operating modes and verify the excess letdown and seal water flows
- 7. Verify ability to maintain steam generator levels and proper operation of feedwater control systems, steam dumps and level instrumentation
- 8. Demonstrate proper functioning of the main steam isolation valves at normal operating temperature and pressure
- 9. Demonstrate the proper operation of steam generator safety I
valves, verifying setpoints with a pressure-assist device and l verifying proper reseating and leakage within specified limits l
l 1
e 75 of 82 l
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F*R B - 18 !
Hl:PS-3 FSAR TABLE 14.2-1 (Cont)
(') 10. Demenstrate the proper operation of precturizer safety and relief valves, and the capability of the pressurizer relief tank to concense a steam discharge from the pressurizer Ma$KT*'
- a. Proptr actuation, operation,Jand response time of the power operated relief valves (PORV) will be demonstrated by simulating a high pressure signal to each valve.2V.5WCT*S
- b. The PORV will be operated manually to confirm valve operability and the ability of the pressurizer relief tank (PRT) to condense a discharge. Leaksge following operation will be verified within acceptable limits.
Discharge header leakage detection instrumentetion will be verified operable in accordance with design requirements.
- d. Safety valve leakage at RCS normal pressure will be verif ted within specified limits. Actual safety valve operation 'will be dentonstrated by hydrostatic bench test to verif y se t points.
- 11. Operate the reactor coolant pumps for a minimum of 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> at .
full flew to achieve approrlmately 1 million vibration cycles on reactor internals. Following hot functional testing, the f]
\-- internals are removed and inspected for vibration effects
- 12. Demonstrate proper operation of reactor coolant pump trips and alarms
- 13. Demonstrate the operability of remote shutdown controls
- 14. Perform or complete those portions of the following system tests (see individual descriptions), which require the RCS to be at or near normal operating temperature and pressures
- a. Reactor coolant system expansion and restraint
- b. Chemical and volume centrol
- c. Boron thermal regeneration
- e. Low pressure safety injection l f. High pressure safety injection
\
- g. Reactor plant sampling l /
l
\_ T/ h. Containment ventilation, I
76 of 82
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^
- 1. Auxiliary building ventilation
- j. Engineered safety features building HVAC
- k. Main steam
- 1. Steam dump control
- m. Steam generator blowdown *
- n. Main feedwater
- o. Steam generator water level control
- r. Reactor plant component cooling
- s. Reactor plant chilled water _
- t. Charging pump cooling
- u. Safety injection pump cooling
- v. Neutron shield tank cooling
- w. Steam generator chemical feed
i
- y. Loose parts monitor system
- z. Reactor coolant and associated system pipang vibration aa. Thermal expansion of piping and components of secondary systems
- 15. Perform or complete tests as necessary to ensure the i
operability of the following systems:
- a. Condensate system
- b. Extraction steam system l
- c. Feedwater heater drains and vents system i d. Turbine plant component cooling system
- e. Turbine plant sampling system 77 of 82
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MHPS-3 p33g TABLE 14.2-1 (Cont)
- f. Normal AC power distribution system
- 16. Perform a controlled plant cooldown, using the steam dump and residual heat removal systems. Demonstrate the capability to de-gas and add hydrogen to the RCS Cr Acceptance criteria i 1
i Systems and components tested will meet specified design, safety l analysis, and Technical Specification requirements.
- l 1
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FSAR Table 14.2-1 Preoperational Test //71 INSERT A
- a. Proper actuation, operation, reset and response time of the valves will be demonstrated. The actuation setpoint and reset pressures of these valves j are a function of instrument calibration.
- b. Operability of instrumentation, controls, interlocks and alarms will be verified.
INSERT B
! The actuation setpoint and reset pressures of these valves are a function of 1 instrument calibration.
i INSERT C
- 17. Demonstrate that the operation of the main steam control valves is
! acceptable within specific limits. Proper actuation and response time of these valves will be demonstrated. The actuation setpoint pressures of
- these valves are a function of instrument calibration.
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T Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-19 Real or Dummy Fuel Assemblies For Vibration Test - Question 640.20(2) (Draf t SER Section 14.2.12)
Question Q640.20:
In FSAR Section 1.8 (Table 1.8N-1, page 6 of 39) the degree of compliance to Regulatory Guide 1.20 (Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing) states that testing and test inspections will be conducted during hot iunctional testing.
- 1. Modify Preoperational Test Number 71 (Integrated Precore Hot Fmctional Testing) Item 11 in FSAR Table 14.2-1 to include a cross-reference to FSAR Section 19.2 for additional information on vibration testing.
- 2. Modify or provide a new startup test description in FSAR Table 14.2-2 that describes the post-core load vibration assessment testing and inspection intended to be accomplished. (Appropriate ref erence may be used for description.)
Response
Refer to revised FSAR Table 14.2-1 for the response to this question.
There is no post-core load vibration assessment testing and inspection intended to be accomplished. Westinghouse stated in WCAP-8780, Verification of Neutron Pad and 17 x 17 Guide Tube Designs by Preoperational Tests on the Trojan I Power Plant, that vibration levels were lower than Indian Point 11 (prototype for Westinghouse 4 loop plant) and were in agreement with predicted results. Also, it was concluded that the internals are free from harmful vibr ations.
Ref er to FSAR Section 3.9N.2.3 f or additional inf ormation.
Additional Concerns identified in Draf t SER:
Technical justification should be provided for not utilizing real or dunmy f uel assemblies in the vibration test in accordance with Regulatory Guide 1.20, Position C.2.2.2.c.
Response
Refer to FSAR Sections 3.9N.2.3 and 3.9N.2.4 for technical justification f or not I using real or dummy f uel assemb!!es in the vibration test. This position has been fomd to be acceptable by the NRC's Mechanical Engineering Branch in their i review of the FSAR as stated in Section 3.9.2.3 of the Millstone Unit No. 3 Draf t i Saf ety Evaluation Report.
1 1
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O Millstone Nuclear Power Station, Unit No. 3 Open Items Procedures and Systems Review Branch PSRB-20 NUREG-0694, item I.G.1 - Question 640.22 (Draf t SER Section 14.2.12)
Question Q640.22:
NUREG-0694, "TM1 Related Requirements for New Operating Licenses," Item I.G.1, requires Applicants to perform "a special low power testing program approved by NRC to be conducted at power levels of greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training." To comply with this requirement, modify Startup Test Number 21 (Natural Circulation) to ensure accomplishment of the following objectives:
Testing - The test should demonstrate the following plant characteristics: length of time required to stabilize natural circulation, core flow distribution, ability to establish and maintain natural circulation with or without onsite and of fsite power, the ability to uniformly borate and cool down to hot shutdown conditions using natural circulation, and subcooling monitor performance.
Training - Each licensed reactor operator (RO or SRO who performs RO or SRO duties, respectively) should participate in the initiation, maintenance, and recovery from natural circulation mode. Operators should be able to recognize when nattral circulation has been stabilized and should be able to control saturation margin, RCS pressure, and heat removal rate without exceeding specified operating limits.
If these tests have been performed at a comparable prototype plant, they need be repeated only to the extent necessary to accomplish the above training objectives. Test data should be used as feedback for simulator verification and update. Attachment 4 to a letter from E. P. Rahe (Westinghouse) to H. R.
Denton (NRC) dated July 8, 1981, contains an acceptable approach for accomplishing the testing objectives hsted above.
. Pcge- AO MNPS-3 FSAR TABLE 14.2-2 (Cont)
- 21. STARTUP TEST - NATURAL CIRCULATION Prerequisites for Testing The low power physics test has been completed. Nuclear steam supply systems and all necessary plant secondary and auxiliary systems are operational. Plant operating procedure prerequisites are met except where special conditions required by this test state otherwise.
Test Objective and Summary The purpose of this test is to demonstrate the plant's capability to remove core heat by natural circulation. The test will be initiated by tripping all reactor coolant pumps and monitoring the establishment of natural circulation,inciJ ; :i:: n _.., s..m 1-..y.. vf ti;; ';; t h :--
p'-* ta stabili n -- ...J .. .. l. .;L. ...., ouu u .c ou.1.sy
_...y t: . int - u ek. ,nn14-- -a.
.3" N O U T Acceptance Criteria Natural circulation cooling can be established and maintained.
23 of 42
f'B RIS -a0 FSAR Table 14.2-2 Startup Test //21 Insert This test will determine the length of time necessary to stabilize natural circulation and will demonstrate the reactor coolant flow distribution by use of incore thermocouples. Effects of changes in charging flow and steam flow on subcooling margin will be determined and subcooling margin monitor performance shall be verified.
This test shall be performed with available licensed reactor operators (RO and SRO) in the control room who will participate in the initiation, maintenance and recovery from natural circulation mode. Data shall also be taken for feedback for the Millstone Unit 3 simulator response to natural circulation. Operators not directly performing the test shall receive training in natural circulation on the Millstone Unit 3 specific simulator with specific instruction in those areas where simulator response may dif fer from actual plant performance.
Specific concerns of Attachment 4 to the July 8,1981 letter from E. P. Rahe to H. R. Denton are addressed as f ollows:
- 1. Manual operation of TDAFW Pump will be performed during Preoperational Test 30, Auxiliary Feedwater. Pre-core hot f unctional testing will verif y Auxiliary Feedwater System capability to maintain SG levels. Since all TDAFW Pump controls are supplied from DC power sources, a loss of AC power verification test will not be performed.
- 2. Pressurizer spray and heater as well as charging and steam flow ef fects on margin to saturation temperature will be tested during Startup Test 2 -
Post-core Hot Functional.
3,4,5. Natural Circulation Test and Station Blackout Test will be performed as Startup Tests 21 and 26, respectively.
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PSRS - SO O +
MMPS-3 FSAR In addition to the normal source range instrumentation, special T submersible neutron detectors are used to monitor flux changes throughout the loading of the core. Data from these instruments will be used to determine, directly or through calculations (i.e., inverse count rate ratio), if an abnormal situation exists. Personnel involved in the monitoring, calculating, or evaluation of data will be briefed on their responsibilities prior to the test.
The procedure starts with the insertion of the temporary nuclear monitor detectors, and those fuel assemblies which contain neutron sources, into the vessel. This is followed by insertion of the remaining fuel assemblies in a sequence to be determined in conjunction with the USSS Vendor. Thrcughout the loading sequence, the following is performed: RCS boron concentration and coolant temperature is recorded; high flux alarms are set at appropriate limits: temporary and source range detectors are monitored visually and at least one channel is monitored audibly; an inverse count rate ratio (ICCR) is calculated. A status board is used to record fuel assembly / detector locations for each step of the procedure. At completion of core loading, a final core configuration is recorded.
Core loading operations are suspended should any of the. following conditions occur
- 1. An unexpected increase in count rate above a specified level
- 3. An unexpected containment radiation monitoring alarm occurs.
- 4. An insufficient number of neutron detector channels becomes available for monitoring.
- 5. ICCR data indicates that an abnormal condition exists.
If core loading has been suspended' for any reason, required surveillances (i.e., boron concentration, water temperature, neutron count rate, etc.) shall continue at the required frequency. Loading operations will not resume until the reason for the suspension has been understood and corrected, or has been evaluated and found a'cceptable.
14.2.10.2 Post-Core Hot Functional After completion of fuel load, the technical specification shutdown margin for a fully loaded core will be verified. Steps are then taken to align and check the operability of instruments, equipment, and control systems necessary for plant heatup. Since the PCHF proceeds in steps from cold conditions to operating temperature and pressure, some prerequisites are not required until just before the appropriate condition for testing. Such prerequisites are identified within the test itself, and met before proceeding. Furthermore, several systems may be tested as appendices of the PCHF. Any 14.2-22 k
- FDRS -a0 MNPS-3 FSAR prerequisites necessary for these tests will be stated in the C applicable test appendix; failure to meet them affects only the test appendix, not the remaining portions of the PCHF. Plant procedures will be used to the maximum extent possible for conducting the PCHF.
Along with general precautions associated with the plant operating procedures , some important precautions for PCHF includes the requirement that reactivity changes be made under the direct supervision of a senior reactor operator, and vigilance to assure any boron dilution does not lower reactor coolant system (RCS) concentration below that required for fueling shutdown.
The PCHF initially prepares the plant for heatup. Upper core internals are installed; the reactor vessel head is placed and the studs are tensioned; cables, ductwork, and insulation are connected; and the missile shield is put in place. While at ambient temperature, the rod control system will be checked out and ro6 drop times are measured. After the prerequisites have been met for plant heatup (RCS filled and vented, reactor coolant pumps (RCP) operable, etc.), the RCS is heated to normal operating temperature and pressure using RCP heat. At selected points in the heatup, RCS leak tests will be performed, operation of instrumentation will be checked and compared, and plant systems will be tested in accordance with the PCHF appendices.
When normal operating temperature and pressure are reached, the following tests will be performed:
(
- 1. Pressurizer . spray and heater effectiveness will be checked.
- 2. RCS design flow will be verified.
- 3. Rod drop times under hot conditions will be checked.
- 4. Flow coastdown will be conducted.
3 n3ER.T >
In addition, items encountered during the pre-core hot functional which were unsatisfactory and systems not previously checked under hot conditions will be tested. This includes a checkout of incore movable detectors, auxiliary feedwater performance verification, and steam dump controls testing.
14.2.10.3 Initial Criticality Upon completion of the PCHF, the primary system is at hot shutdown with reactor coolant pumps operating, RCS temperature controlled using the steam bypass / dump system, and RCS boron concentration equal to or greater than the value for core loading. Remaining deficiencies are reviewed by the JTG and resolution obtained prior to authorization of performance of the initial criticality procedure.
In addition to the regular plant' systems necessary for initial criticality, special equipment, such as a reactivity computer and recorders for monitoring / plotting data, are checked out and verified as operational.
14.2-23
PSRB-ao 1
(FSAR pg.14.2-23) '
Insert
- 5. Effect of pressurizer heaters and normal spray and changes in charging and steam flow on margin to saturation temperattre will be checked. Auxiliary spray will not be tested due to cyclic limitations based on dif ferential tem perat ure, k _ __
O .
Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSRB-21 Regulatory Guide 1.68, Rev. 2, Appendix A -
Question 640.26 (Draf t SER Section 14.2.12)
Ques tion Q640.26:
Our review of your test program description concludes that the operability of several of the systems and componcats listed in Regulatory Guide 1.68 (Revision
- 2) Appendix A may not be adequately demonstrated by your initial test program.
Expand FSAR 14.2.12 to address the following items:
NOTE: Although some of these systems are designated for testing in Preoperational Test Number 71 (Integrated Precore Hot Functional Testing) Part 15, individual test descriptions f or these systems should be included in FSAR Chapter 14 to adequately describe what testing will be done. Inclusion of a test description in FSAR Chapter 14 does not necessarily imply that the test becomes subject to FSAR Chapter 17 Quality Assurance Program cont rols. Certain tests to be performed prior to fuel loading to verif y system operability may be referred to as " acceptance tests" to distinguish them from "preoperational tests" subject to FSAR Chapter 17 test control.
, Preoperational Testing R.G .1.68 FSAR Appendix A Section Description 1.a(2)(f) 5.4.12 Loop stop valves 1.a(2)(h) 5.4.15 Reactor vessel head vent system 1.d(9) 10.4.7 Condensate storage system f.e(5) 10.4.7 Steam extraction systern 1.e(8) 10.4.7 Condensate system f.e(10) 10.4.7 Feedwater heater and drain systems 1.3(12) 10.4.2 Condenser air evacuation system 1.g(l) 8. 3.1.1.1 Normal ac power distribution system 1.h(5) 7.6.6 Reactor coolant system loop isolation valve interlocks
1 Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSRB-21 Continued i R.G.1.68 FSAR l Appendix A Section Description 1.h(3) 6. 3. 5 Ref ueling water storage tank level and temperature indication l
1.h(10) 9.2.5 Ultirnate heat sink l
l 1.J(7) 6. 3.2.5 Leak detection systerns used to i detect f ailures in ECCS and l containment recirctdation spray
! systems located outside containment 1.j(16) 10.4.7 flotwell level control systerns 1.J(17) 10.4.7 Feedwater heater temperature, course of postulated accidents:
a) containment wide range pressure Indicators b) containment sump level inonitors c) containment radiation monitors d) hurnidity monitors 1.J(24) 7.1.1.5 Reactor control and ESF annunciators 1.k(2) 12.5 Personnel monitors and radiation survey instrument tests 1.k(3) 12.3 Laboratory equipinent used to analyze or measure radiation levels and radioactivlty concentration i
1.k(4) 6.5.1.4 IlEPA filter and charcoal adsorber ef ficiency and inplace leak tests. Modif y the appropriate test abstracts to ensure that testing in accordance with Regulatory Guide 1.52 (Design, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light Water-Cooled Nuclear Power Plants),
l Positions C.S.a - C.S d, and Regulatory Guide l.140 (Design, Testing, and Maintenance Criteria for Normal Ventilation Exhatat System Air l
Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems lleview Branch PSRB-21 Continued R.G.1.68 FSAll Appendix A Section Description Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants),
Positions C.S.a - C.S.d, is accornplished.
- 1. l(8) 9. 3. 2 Turbine plant sampling system 1.m(3) 9.1.4 Operability and leak tests of sectionalizing devices and drains, and leak tests of gaskets or bellows in the ref ueling canal and f uel storage pool 1.m(4) 9.1.4 Dynamic (100%) and static (125%) tests of cranes, holsts, and associated iuel storage and handling systems 1.n(3) 9. 2.7 Turbine plant component cooling systern 1.n(16) 6. 3.2.2.2 Cooling and heating systems Ior tank 1.n(18) lleat tracing and f reeze prutection systems 1.o(l) 9.1.$ Polar crane dynarnic (100%) and static (12$%)
loading tests Power Ascension Tests S.w Containment penetration coolers. Provide a preoperational test description or, on those penetrations where coolers are not used, provide a startup test description that will demormtrate that concrete ternperatures surrounding hot penetrattora do not exceed design limits.
3.11 13.3.2 Demonstrate that the dynamic resporse of the plant is in accordance with design for limiting reactor coolant pump trips. The method Ior initiating pump trip should result in the iantest credible coastdown in flow.
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k Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch i P5RB-21 Continued [
Response
Refer to the discussion below and to revised F5AR Tables 14.2 1 and 14.2 2 for r the response to this question.
Instrumentation testing is performed within the phase 1/2 testing of the system to which the instruments belong. Instrument loop continuity and calibrations are i perforrned during this testing. Indications, alarm and comptuer readings are l Verified where appropriate. This includes annunciation. t Ref er to FSAR Table 1.81 for conformance to Regulatory Guides 1.52 and 1.140. ;
Tests and calibrations ior personnel monitors, radiation survey instruments and l laboratory equiprnent are performed in accordance with station procedures which l are currently in use on alte for Units 1 and 2. '
Tabic Q640.26 1 Indicates where each test abstract may be found.
Additional Concerns identitled in Draf t 511R 1.g(!) - Testing should be provided to verif y equipment operability at maximum '
and minimuni design voltage (BTI' P5B 1, paragraph B.4).
l.m(4),1.0(l) - Documentation should be provided which ensures that the !
construction load testirg of the polar crane and crancs, holsts, and associated ;
f100%
uel storag(e load dynamic and test).
handling system is accornplished at 125% load (static test) and .
3.w - llot penetrations which are not serviced by reactor plant component '
l cooling should also be monitored during startup testing.
i-Responses t
.l_40)
Testing will be performed as specified in response to NRC Question Q430.ll. !
l.in(4h1.0(l) ;
3 l Ref er to revised if5AR Table 14.2 1, Preoperational Test 4.
Crane load testing of the polar crane and cranes, holsts, and associated fuel t storage and handling system is accomplished at 123% load (static test) and 100%
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Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems lleview Branch PSitn-21 Continued load (dynamic test) by our orchitect-engineer, Stone and Webster Engineering Corporation during construction. Crane test documentation is obtained during equipment turnover to the NNECO organization and is attached to the applicable Phase I test procedure document. Subsequent testing will be in accordance with Technical Specification requirements. >
i liefer to revised F5All Table 14.2-2, Startup Test 42.
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!!NPS-3 FSAR TABl.E 14.2-2 (Cont)
- 42. STARTUP TEST - CONTAIll!!ENT PENETRATIO!! TE!!TERATURE M0!!!TORIllG -
Prerequisites for Testing The plant is at approximately 30, 50, 75, 90, and 100 percent power.
T_est Objective and Summary Testing will monitor the temperature of hot penetrations serviced by 640.26 reactor plant component cooling. 4 Acceptance Criteria Luh.7 A d -4 Reactor plant component cooling can maintain the penetrations within design temperature limits.
t
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Amend. ment 5 42b of 42 November 1903
PORB -a t (FSAR Table 14.2 2)
Startup Test //42 Insert A Additionally, testing will monitor the temperature of other penetrations I determined to be hot penetrations but not serviced by reactor plant component l cooling.
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PSRb "3 \
liti?S-3 I6bR e
TABLE 14.2-1 (Cont)
- 4. PREOPERATIO!!AL TEST - POLAR CRANE Prerequisites for Testing N as _
General prerequisites have been met, el; - c----t :;;ti , ..~.~...v a- *
- - -..-Hnn load test b'e $---- -- l1;;,l. -
Test Objective and Summary This test will verify operability of polar crane control circuits and ability to handle the reactor vessel head and various internals components. l640.27 Acceptance Criteria The crane control circuits and interlocks function in accordance with design. The crane is capable of installation and removal of the reactor vessel head and those internal components placed during cold hydrostatic and hot functional testing.
L Amendment 5 7 of 82 !!ovember 1983
. p5RB -al D
(FSAR Table 14.2-1) ,
Preoperational Test 44 t
insert B All component testing including the construction 123% static and 100% dynamic load tests have been completed.
= 0 0 Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch PSRB-22 Preoperational Tests 76 Question 640.27 (Draf t Sell Section 14.2.12)
Question Q640.27 Modify FSAR Figure 14.2 5 to include the following preoperational tests listed in FSAR Table 14.2-1.
4 -
Polar Crane 10 - Residuallicat Removal 14 -
Reactor Plant Sampling 26 - Steam Dump Control 27 -
Steam Generator Blowdown 39 -
Radioactive Liquid 4 aste 53 -
Reserve Station Service Transformers 54 - Communications l
65 - Emergency Lighting 72 -
Reactor Coolant and Associate System Expansion and Restraint 73 -
Reactor Coolant and Selected Systems Piping Vibration
, 74 -
Thermal Expansion of Piping and Components of Secondary
! Systems 75 -
Control System Test for Turbine Runback Operation l
Response
Ref er to revised FSAR Figure 14.2-3 for the response to this question.
Additional Concerns identitled in Draf t SER:
Preoperational tests 76 through 84 should be included in FSAR Figure 14.2-5.
Responsen Ref er to revised FSAR Figure 14.2 5.
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U^** W.m (,,4 w o w u. A,at FIGURE 14 2 *S PREOPERATIONAL TEST PHASE MILLSTONE NUCLC AR POWER STAfl0N UNIT 3 C FINAL SAFtf Y ANALYSIS REPORT AWtNDMLNT& Novttsein even
Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systerns Iteview !) ranch I P51 tit-23 Swing Load Test - Question 640.28 (Draf t S!!!t Section 14.2.12)
Certain startup test listed below do not specif y the power level at which the test l will be conducted,instead stating that testing will be conducted at selected or various power levels. Modif y the individual test abstracts to include the specific ;
mwer level values at which each of the tests will be conducted. Modify F5All i rigure 14.2 6 to indicate which tests will be conducted during each power !
plateau during the startup prograin, or provide a clarification stating that these ,
tests will be conducted at power levels consistent with Regulatory Guide 1.68, :
llevision 2. l l 1 l 14 -
Loose Parts Monitoring Systern l l i
- 13 - Water Chemistry Control
! 16 - Itadiation 5urvey ,
23 - Operational Alignrnent of Nuclear Instrurnentation i 29 -
Process and Elfluent Radiation Monitori.'s; Systern i 30 - Core Periortnance i 31 -
Power Coef ficient Measureinents 33 - Ventilation System Operability 34 -
Turbine Generator and Feedwater Turbine Operability Test ;
33 -
Calibration of $ team and Feedwater Flow Instrumentation at Power '
37 - Load Swing Test !
Response
Ref er to revised F5AR Table 14.2 2 Ior the response to this question. '
Additional Concerns identitled in Dralt 5tR The load swing test (startup test 37) should include testing at 30% power in accordance with Regulatory Guide 1.68, Appendix A,5ection 3.h.h. <
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Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systerns iteview Dranch P5ftlh23 (Cont.)
llesponse Load Swing Test (Startup Test 37) will not be perforrned at 30% power for the
!ollowing reasons:
- 1. Load 5 wing Test will be perfortned at 30%,73%, and 100% plateaus.
- 2. A 4% Load swing Test will be perforrned at 30% as part of the power coef ficient test (5tartup Test 31).
- 3. % e will conduct a f ull reactor trip at 30% which is significantly more lirniting than the 10% load swing test.
Itefer to revised F5 Alt section 14.2.7.7 for itegulatory Guide 1.68, llevlilon 2 conforinance Staternent Appendix A, Section 3.h.h.
b q
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' 4 PSRt3 - JL3 !
., c HNPS-3 FSAR i l
l 14.2.7.4 Regulatory Guide 1.37, Rcvision 0 - Quality Assurance C>
I Requirements for Cleaning of Fluid Systems and Associated 640.1 l components of Water-cooled Nuclear Power Plants The Hillstone 3 initial test program will conform to the intent of Regulatory Guide 1.37.
14.2.7.5 Regulatory Guide 1.41 Revision 0 - Preoperational Testing of Redundant Onsite Electrical Power Systems to Verify Proper Load Group Assignments For position on Regulatory Guide 1.41. see FSAR Section 1.8.
~
14.2.7.'6 Regulatory Guide 1.52,:,, Revision 2 - Design, Testing, and Maintenance Criteria for Post Accident Engineered Safety Feature Atmosphere cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Huclear Power Plants For position on Regulatory Guide 1.52, see FSAR Section 1.8. .
14.2.7.7 Regulatory Guide 1.68, Revision 2 - Initial Test Programs for Water Cooled Hucleat Power Plants The Hillstone 3 initial test program will conform to Regulatory Guide 1.68, except as specified in this sections ,
O -- O
$5..$- -=.,_-....,,..,..-.....-...A
$ U N f 9 5 N f'i2f_ h b ,'" =
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( -&. During power escalation, testing will be conducted at the 30 percent power level instead of at the 25-percent power level. Westinghouse supplied plants have generic data for the 30 percent level which they do not have at the ,'
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T*4% %."C' D -b 25 percent level (Section C.8 Appendix A, Section 5). '
$ 4. The H5tv closure test will be performed at less than '1i 20 percent power to demonstrate,the proper dynamic response Og, 1
of the plant and to verify proper integrated operation of l plant equipment. Plant response to a full power trip will be verified by the generator ' trip at 100 percent power.
- Closure of the HS!Vs at 100-percent power would not provide any additional information significant enough to warrant
, subjecting the plant to such a severe thermal transient 3
-(AppendiN A, Section 5.m.m). ,
in i 4 =A The loss of feedwater heaters test will not be performed.
9 '
, since plant response to load swings and large load '
reductions is demonstrated in other tests, there is no need to subject the plant to this additional transient (Appendix A, Section 5.k.k).
( 4,-
H111 stone 3 does not have a partial scram feature (Appendin A, Section 5.j). f Amendment S 14.2=17 November 1983 ,
a : ':c ' .
, ,, a u h,u n ~ u 6 $ E '
1 e:nte-as ;
[
! (FSAR pg.14.2-17)
J l Insert B l
i 2. Load swing testing will be conducted at the 30%,75%, and 100% plateats. '
j A 4% load swing test will be conducted at the 50% plateau as a part of the j power coefficient test (see FSAR Table 14.2-2 Startup Test 31). ,
j Additionally, a f ull reactor trip at the 50% plateau will be conducted.
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J..._______.____._____.__._ _ _ . _ _ _ _ _ _ _ _ _ -
. P3R.6 -5Li HilPS-3 FSAR yA Demonstration of the design capability of reactor residual (',
or decay heat removal systems will be done during power N. <
ascension testing caly if it is not done during hot functional or low pouer tests (Appendix A, Section 5.1).
] ~7-- The following systems will be tested during the startup test phase only if they are not completed during the preoperational test phase:
Reg. Guide Section Component Tested Appendix A, Section 4p Pressurizer and main steam relief valves Appendix A, Section 4r Reactor coolant puri-
, fication and cleanup system ,
Appendix A, Section 5.c.c Gaseous and liquid waste radioactive waste systems
% -h- Portions of Appendix A, Section P,.s, will not be conducted.
Hi11 stone 3 does not have an integrated control system or a reacter coolan; flow control system.
g A- The auxiliary (startup) and emergency feedwater control
)
systems and the steam pressure control systems will be -
tested before the power e.scension test phase since these systems arr not used at power levels above the low power 1
operstion modes (Appendix A, Secticn 5.4).
(O Mr The individual rod position indication system is the primary l
l means for determining control rod micalignments. The design of the nuclear instrumentation is not intended to detect a misaligned control rod but rather to detect anomalous core -
conditions. Therefore, testo will not be conducted in accordance with Appendix A, Section 5.1. Ilowever, data on nuclear instramentation characteristics will be obtained durinJ the core performance test.
( \ .H r When several emergency loads are identical in type, size, and manufacturer, only one of these loads will be started and operated with the reaximum and minimum design voltage available. Testing every emergen:y load would not provide any additional assurances significant enough to warrant the required increase in the scope of the test program (Appendix A, Section 19 ).
14,2-18
, , . . ..s
Millstone Nuclear Power Station, Unit No. 3 Open items Procedures and Systems Review Branch Question Q640.15 (Section 14.2.12)
In accordance with the test requirements listed in Regulatory Guide 1.41 (Preoperational Testing of Redundant Onsite Electric Power Systems to Verify Proper Load Group Assignments), Position C.2:
- 1. Modify Preoperational Test Number 50 (125 V dc Distribution) to incorporate testing to verify that at the minimum and maximum design battery voltages, required Class lE systems can be started and operated. At minimum battery voltage, with charges deenergized demonstrate capability to start all IE loads. Then, with the chargers energized verify ability of the chargers to supply loads and charge bat teries. For more information on problems with maximum battery voltage conditions, see IE Information Notice 83-08, March 9,1983.
- 2. Modify Preoperational Test Number 53 (Reserve Station Service Transformers) to demonstrate the proper operation of transformer cooling under design load or describe how data from testing under available load will be extrapolated to verify cooling capability under design loading.
Response
} Refer to revised FSAR Table 14.2-1 for the response to this question.
Preoperational Test Number 30, 125 V de distribution, tests the design
, capabilities of the 125 V de system. The concerns of Regulatory Guide 1.41, i
Position C.2, are met in Preoperational Test Number 67 - Engineered Safety i Feature Test with Loss of Normal Power.
l By design, the components powered from the 125 V de system will function between the maximum and minimum bus voltages. Refer to the response to Question 430.41. Since the de components are designed for minimum expected voltage, the testing of all IE loads for minimum voltage starting is not necessary. The cooling capabilities of the reserve station service transformers
- (RSST) have been demonstrated by vendor test. With the RSST under rated load, equilibrium temperatures were below design limits.
Additional Concerns identified at Maren 21,1984 Meeting:
Response should reference testing accomplished to address the concerns of item 430.!! (conformance with Branch Technical Position PSB-1, NUREG-0800, Appendix 8A). Additionally, the response to item 640.15 should be modified 3
accordingly.
Response
j Refer to revised FSAR Table 14.2-2, Startup Test 26.
l~ Ovlo. Is
}NPS-3 FSAR- .
TABLE 14.2-2 (Cont)
- 26. STARTUP TEST - STATION BLACKOUT Prerequisites for Testing The plant is in the 10 to 20 percent power range with all plant loads being supplied by the Millstene 3 generator.
Test Objective and Summary Tnis test will demonstrate that the plant responds as designed following a plant trip with no offsite power. The reactor will be tripped. The diesel start, load sequencing, and plant response including natural circulation will be monitored. The turbine-driven auxiliary feedwater pump shall be run for a minimum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with motor-drit en auxiliary 640*12 feedwater pumps and turbine-driven auxiliary. feedwater pump cubicle ventilation secured.
Acceptance Criteria
'. I.R5ERT" The plant responds > in accordance with design. The turbine-driven 640.12 auxiliary feedwater pump will remain within design limits and pump room ambient conditions do not exceed environmental qualification limits for safety related equipment in the room. >
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FSAR Table 14.2-2
- Startup Test 26 4
INSERT AC power to the inverters and battery chargers will be removed for a period of two hours to force battery operation.
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Millstone Nuclear Power Station, Unit No. 3 i Open Items i Procedures and Systems Review Branch Question Q640.16 (Section 14.2.12)
- 1. In accordance with Regulatory Guide 1.108 (Periodic Testing of Diesel Generator Units used as Onsite Electric Power Systems at Nuclear Power
- Plants), Position C.2.a.4, modify Preoperational Test Number 51 (Diesel Gene'rator) or Number 67 (Engineered Safety Features Test With Loss of Normal Power) to demonstrate proper operation during diesel generator load shedding, including a test of the loss of the largest single load and complete loss of load and verify that the voltage requirements are met and that the overspeed limits are not exceeded. Your testing should, in addition, provide assurance that any time delays in the diesel generator's 1
restart circuitry will not cause the supply of starting air to be consumed in
, the presence of a safety injection signal (see I&E Information Notice 2
Number 83-17, March 31,1983).
- 2. Modify Preoperational Test Number 51 (Diesel Generator) .to include
- testing to ensure the satisfactory operability of all check valves in the flow path of cooling water for the diesel generators from the intake to the discharge (see I&E Bulletin No. 83-03
- Check Valve Failures in Raw Water Cooling Systems of Diesel Generators).
Response
Refer to revised FSAR Table 14.2-1 for the response to this question.
There are no check valves in the service water lines that provide cooling water to the diesels. As such, the concerns of I&E Bulletin 83-03 do not apply to the Millstone Unit No. 3 design.
Additional Concerns identified at March 21,1984 Meeting:
, Modify Preoperational Test Number 51 (Diesel Generator), Test Objective and i
Summary Item 7, in accordance with the revised clarification in FSAR Table 1.8-1 to Regulatory Guide 1.108 (Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants), Position C.2.a.4, as revised by the response to item 430.16. The test of the loss of the single largest load and complete loss of load should be conducted with the diesel initially at its 2000-hour rating. '
Response
Refer to revised FSAR Table -1.8-1, Regulatory Guide 1.108.
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MNPS-3 FSAR TABLE 1.8-1 (Cont)
R.C. FSAR Section Ho t 1i119 Degree of Complicance Reference 1.108 Periodic Testing or Diesel Comply, with the following clarirications and 8.3.1 Cenerator Units Used as Onsite exceptions:
Electric Power Systems at Nuclear Power Plants Section C.2(a)2: Proper operation for design-(Rev. 1, August 1977) accident-loading-sequence will be demonstrated under conditions as close to design as possible.
Section . (a)3: . full-load rrying c . ill will demon ted for -hour p od at le 430.15
? -hour ing.
Sectio .2(a)4- he tee or the i s of th ([) __
lar et singi oad of comp .e loss I will be emont .ra ted wi the uni pe ra- li 430*16 ino at tt 2000-h or rat in Section C.2(a)9: Comply as stated in the ERRATA dated September 1977.
1.109 Calculation of Annual Doses Comply 13.3.1 to Man from Routine Releases of Reactor Erriuents for the Purpose of Evaluating Compliance with 10 CFR Pa rt 50, Appendix I .
(Rev.1, October 1977) 1.110 Cost-Benefit Analysis for Comply Radwaste Systems for Light-Water-Cooled fluclear Power Reactors
( Rev. O, Ma rch 1976) .
1.111 Methods for Estimating Atmos- Comply 2.3.5.2.3 pheric Transport and Dispersion of Caseous Erriuents in Routine Releases f rom Light-Water-Cooled Reactors (Rev.1, July 1977) t w
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Amendment 3 39 ors 58 August 1983