ML20058F509
| ML20058F509 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/24/1988 |
| From: | NORTHEAST NUCLEAR ENERGY CO. |
| To: | |
| Shared Package | |
| ML19311B205 | List: |
| References | |
| NEO-2.25, NUDOCS 9312080157 | |
| Download: ML20058F509 (75) | |
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NUCELAR EN'!NEER:NG AND OPEF.ATIONS FROCEDURE FE0 2.25 IDENTITICATICN A'iD IMPLEMENTATION OT NRC REPORTING REQUIREMDTTS (10CTF.50.7I, 200TF.50.73, AND 100TR50.9)
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CONCUP.RENCE
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Respons2 ele Indavacual CONCURRENCE 1~
C Director, Quality Serv ces Department APPROVED y s.: S
[-/IN Serriq(' Vice Prpeent Date Nuclear Engineering and Operations RIVISION 1
ETTECTIVE DATE OCT 2 41998 9312080157 931130 PDR ADOCK 0500024E Q
PDR a
e NUCLEAR ENGINEEP.INO AND CPERATIONS PROCEDURE NED :.25 IOEN!!TICATION AND IMPLEMENTA!!ON OF NRC REPORT:N REQUIREMENTS (10CTR50.72, 100TR50.73, AND 10CTR50.9) 1.0 T'.'RPO SE To provide instructions to enable Nuclear Engineering and Operations (NED) personnel to identify:
1.1 Potential Reportable Items to be evaluated for reporting to the NRC.
1.2 The NRC notification and reporting rules to be implemented in the evaluation and reporting of a Reportable Item, 1.3 The appropriate personnel and procedures necessary for the implementation of the appropriate notification and reporting rule (s).
Lover tier procedures may be used to implement this procedure.
2.0 APPLICABILITY This procedure applies to the NEO Group, which includes the Northeast Nuclear Energy Ccapany (NNECO) and the Connecticut Yankee Atomic Power Company (CYAPCO).
This is an NEO Quality Procedure.
3.0 RETERENCES 3.1 Source Documents 3.1.1 10CTR50.72 "Immediate Notification Requirements for Operating Nuclear Power Reactors."
3.1.2 10CTR50.73
" Licensee Event Report System."
3.1.3 10CTR21
" Reporting of Defects and Noncompliance."
10CTR50.9
" Completeness and Accuracy of Informa-h 3.1.4 tion."
NEO 2.25 Rev. 1 Page 1 of 10
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2.2 Suceertsnr Documents 3.2.1 NEO 2.01 - ~!s;lementation of the Requirements of Part 21 of Title IC. Code of Federal Regulations:
Reporting of Oefects anc Noncompliance."
2.2.2 Tacility operating licenses (i.e., License Condi-i tiens. Crders Modifying License. Exemption's, Safety Techn1 cal Spec 2ficat2ons.and Environmental Protection Plan (Millstone 3)) for each NU Operating Nuclear Generating Unit.
'i 3.2.3 10CTR50.49
- Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants."
3.2.4 NED 2.13
- Management of Nuclear Power Plant i
Records."
3.2.5 NUREC-1022
" Licensee Event Recort System" and Supplements No.1 (February 1984) and No. 2 (September 1985).
t 3.2.6 Generic Letter 86-15 "Information Relating to j
Compliance with 10CTR50.49," (September 22. 1986).
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3.2.7 Generic Letter 88-07 "Information Relating to 3
Compliance with 10CTR50.49," (April 7, 1988).
j 4.0 DETINITIONS 4.1 Acronyms Used In This NEO Procedure-Code of Federal Regulations CTR CYAPCO - Connecticut Yankee Atomic Power Company GTL
- Generation Tacilities Licensing Justification for Continued Operation l
JC0 LER Licensee Event Report NE0
- Nuclear Engineering and Operations
-l Northeast Nuclear Energy Company NNECO i
i Nuclear Plant Records Tacility
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NPRT Nuclear Regulatory Commission l
NRC Northeast Utilities NU Nuclear Regulation i
NUREG NUSCO Northeast Utilities Service Company Plant Incident /Information Report PIR Reportability Evaluation Tors (Tigure 7.2) l RET SSH Substant2a1 Safety Eazard NEO 2.25 Rev. 1 Page 2 of 10
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4.2 Peeert/Netificatsen - A vercal. vratten. or-electronically transma t tee commun2 cation to the NRC Staf f inf orming tnem of i
a Reportacle Ite. in conformance with tne requ:rements in 100TR50.71.10 TR30.73, or 100TR50.9(b).
4.3 Feoortsele Itee - An event. incident, condition, deviation, cetsctancy, cetect, noncompliance, or act that could affect i
nuclear plant safety and must be reported to the KRC in I
accorcance vath 100TR50.72. 100TR50.73, or 10CTR50.)(b) and lover t:er report 2ng proceoures.
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4.4 Reoerting Rules - The reporting rules in 100TR50.72.
/ \\l 1DCTE50.73, ano 10 TR50.9 that require all potential report &ble items to be first evaluated for reportability, ar.d reported in confermance with the requirements of J;
10CTR50.72, 10CTR50.73, or 10CTR50.9(b) and/or applicable station procedures.
4.5 Substantial Safetv Hazard Reecrt - A report to the NRC of an SSH 11 cetermanec necessary 2n conformance with 10CTR21 and l
NEO 2.01.
i 4.6 Justification fer continued Operation - Vritten justifica-3 2
tien as to vny a paant may sately operate with a known deficiency in analysis, design, or operation of a structure, system, or component. Also there can be no major reduction in the degree of protection provided to the public health and safety.
4.7 Significant Imolication - An issue that is determined a
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reportacle stem uncer 10CTR50.9(b) which has a major impact on public health and safety and is not oth'ervise required to i
be reported to the NRC under other reporting rules.
5.0 RESPONSIBILITII5 5.1 Fanager. Generation Tacilities Licensing i
5.1.1 Provides reporting guidance to NEO personnel, as l
necessary.
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5.1.2 Maintains a sequentially numbered log and file of potential reportable items.
5.1.3 Designates Managers / Supervisors to perform report-l l
ability evaluations.
5.1.4 Coordinates the preliminary notification to the Superintendent when the technical evaluation of an RET requires more than 30 days.
NEO 2.25 Rev. 1
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Page 3 of 10
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5.1.5 Prepares a re em=eneatica on the reportaoility of j
inose potent.a. reportaole items for vnich an RIT has eeen initiates ano fervares that recommenca-tien tanc any JCOs) to the appropr2 ate Superin-tendent.. Comoleted JCDs snall also be transtatted 3.
to toe NR3/SN7Ji for review.
5.1.6 Maintains tais procedure current.
5.2 Su=erintendent
'Jni t Sucerintendent. Station Suoer2ntencent. Statsen serv 2ces Suoer2ntencent at KNECO.
CYAPCO. Or Betn 5.2.1 Determines the reportability of potential j
reportacle items in accordance with the requirements of 10CTR50.72,-10CTR50.73, and 10CTR50.9(b) and/or upplicable station procedures.
5.2.2 Notifies the NRC of reportable items in accordance with the requirements of 10CTR50.72, 10CTR50.73.
or 10CTR50.9(b) and/or applicable station procedures.
5.2.3 Determines whether potential reportable items.
should be dispositioned via a PIR or through an RET and if a JC0 is needed.
5.2.4 Ensures that a PIR or RET is initiated for any circumstance that results in the need for pre-3_
paring a JC0 at the site.
5.3 NUSCO Manarers/ Supervisors 5.3.1 Complete technical evaluations of potential jff l
reportable items within the time'specified on the RET.
i 5.3.2 Provide recommendations on the reportability of those potential reportable items for which an RET has been initiated.
5.3.3 Provide a preliminary recommendation and schedule j
to the Manager, GTL, when the technical evaluation j vill take more than 20 days.
5.3.4 If a JC0 is required, provide a separately j
j(hg j
documented JC0 that references the RET.
o 5.3.5 Ensure an P.ET is initiated for any circumstance j
that results in the need for preparing a JC0 at NUSCO.
NED 2.25 Rev. 1 Page 4 of 10 i
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i 1.4 Nuclear Review Beare!!!!e Nuclear Peview bearc (h'RE/SW E) f Reviews esta Reportacility Evaluatsen vnich involves a JCO.
Rev2evs the JC0 anc the assoc 2ateo documents at the next regularly seneduled NRB/SNR3 zeeting.
6.0 INSTRU C ONS L
The secuence of pratary actions, and the personnel responsible fer tnem, are shovn in Tigure 7.1 - Flovenart.
t Suggested maximum processing times are specified on the RIT.
6.1 Section 1.0. Ordrinater and Suoervisor 6.1.1 NUSCO, NNECO, and CTAPCO personnel, upon becoming aware of a circumstance that results in the need h for a reportable item, or a potential significant 10CTR50.9(b) and determine the implication or JCO, shsll review the ites against the requirements of 10CTR50.72, 10CTR50.73, and 3
I tien and possibly a subsequent prompt NRC notifi-cation or 1.ER in accordance with the steps listed below.
Redundant evaluation and reporting is not required. Thus, a significant implication vould not be evaluated or reported as such if it has already been determined that the item is report-able to the NRC under 10CTR50.72,10CTR50.73, 10CTR21, or other reporting requirements.
6.1.1.1 NNECO and CTAPCO personnel (hereinafter called Originator), with knowledge of a potential significant implication or reportable iten, shall promptly inform the superintendent verbally or via a i
FIR, in accordance with applicable sta-l tion procedures.
If NUSCO engineering support is needed to evaluate a potential significant h
implication or reportable ites, the Superintendent may instruct their personnel to initiate an RET and to disposition the item in accordance with this procedure.
6.1.1.2 NUSCO personnel (hereinaf ter-called Originator), identifying a potential significant implication or reportable b
item, shall initiate an RET (Tigure 7.2) ~~
NIO 2.25 Rev. 1 Page 5 of 10
j anc forvaro it to their Supervisor. The Supervssor snall confirm that the orig:-
nator has clearly identified the poten-tial safety icplication before forvaro-ing to the Manager, CTL. The Supervisor I
need not concur with the Originator's decision to initiate an RIT.
6.1.2 The Originator of an FIT shall complete Section 1.0 of the FIT (Tigure 7.2):
i 6.1.2.1 Describe the potential significant
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implication or reportable item. Attaen j
any required additional documentation.
6.1.2.2 Describe the potential significant A
implication and safety implications.
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6.1.2.3 Identify the sectiens of 10CTR$0.72 10CTR50.73, or 10CTR50.9(b) that poten-
,2 tially require the item to be reported.
6.2 Sectien 2.0. Manager. GTL 6.2.1 Upon receipt of the RIT, the Manager, GTL, shall l
assign a log number and log in the RET.
I 6.2.2 The Manager, GTL, shall contact the Superintendent vho shall determine if a NUSCO reportability eval-untion is needed.
6.2.3 If a NUSCO reportability evaluation is not deemed necessary, the Superintendent shall ensure that the item is evaluated in accordance with applic-1 able station procedures.
6.2.4 If the Superintendent has determined that the A
potential significant implication or reportable S1 item vill be evaluated according to applicable station procedures, the Manager, GTL, shall file i
the RET, note the dispositica in the log, and j
inform the Originator of this disposition. No i
further NUSCO action is required.
6.2.5 If the Superintendent has determined that the potential significant implication or reportable ites requires further NUSCO evaluation to deter-aine reportability, the Manager, GFL, shall desig-nate the Supervisor (s) or Manager (s) who vill be responsible to perform the evaluations and forvards the RET to the first designated j
Supervisor / Manager. Consideration vill be given j
l NE0 2.25 Rev. 1 Fage 6 of 10 l
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D to the need f er a multidiscipline review and ass 2gnments mace concurrently.
6.2.6 If the Superantencent has determined that e JC0 is required, then a JC0 shall be separately prepared anc provided to the Manager, CTL. for transmittal to the Superantencent and the NRB/ShTdi for review.
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6.3 Section 3.0. Technical Surervisor/Mansrer i
6.3.1 Each designated Supervisor / Manager snall review I
the potent 2al significant implication or report-able item to determine the potential safety impact. A root cause analysis, if needed, can be l
provided separately from the RIT. Each designated Supervisor / Manager shall recommend corrective t
action, if appropriate, and provide a reportabil-ity recommendation within 20 calendar days from i
the time the Originator initiates the RET form 1
j (Step 6.1).
If they determined the ites to be reportable, they vill identify the applicable i
reporting paragraphs of 10CTR50.72, 10CTR50.73, or i
h 10CTR50.9(b).
I 6.3.2 If Step 6.3.1 cannot be completed within d
20 calendar days from the time the Originator initiates the RET form (Step 6.1), then the Supervisors / Managers shall provide a preliminary i
evaluation of reportability using engineering i
judgment to the Manager, CTL, by the end of the 20-day schedule. The preliminary evaluation shall also contain an estimate of the schedule for-com-i plating a final reportability recommendation.
i 6.3.3 If a JC0 appears necessary, obtain the concurrence g of the Superintendent through the Manager, GTL, that a JC0 is needed. The Supervisors / Managers
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performing the JC0 shall separately provide it to the Manager, GTL as a separate document, after review / approval by the discipline management. The Manager, GFL, assures that multidiscipline tech-nical reviews result in an integrated JCo. vhen appropriate. The minimum signoff level for a JC0 is a sanager.
Normally, the need for a JC0 vill be decided by
'i the Superintendent at the beginning of a Report-ability Evaluation (step 6.2.6).
Bovever, any l
other issue that generates the need for a JC0 i
shall first (if possible) be evaluated for report-ability under NED 2.25 or station procedures. A NEO 2.25 Rev. 1 Page 7 of 10 f
JC0 may or may net te revievec by the NRC or
_,...... m...,..., m,..........
vould neee to be a :ressec in a JC0 on equipment qualificat2cn (for example) are prov2ded in l
10CTR50.49(1) anc Gener:c Letters 86-15 and 88-07.
7 The thougnt process for tne JC0 should be similar to that usec in the performance of a safety evaluation (NEO 3.12)' to the extent applicable s2nce it deterrines acceptacility regarding puolic health anc safety.
The seneeule fer ccepleting the JC0 shall not j
icpact the senedule-f or completing the PIT. If the senedule for' preparing a JC0 requires the use of engineering judgment, then the JC0 may be subsequently updated using standard engineering practices.
6.4 Section 4.0. Manarer. CTL l
6.4.1 The Manager, CTL, shall review the PIT and attachec documentation and provide a recem-mencation as to whether the item is reportable 1
vithin 30 calendar days from the time the Originator initiates the PIT form (Step 6.1).
6.4.2 If the item appears to be reportable, the Manager, GTL, shall prov2de early notification to the Superintendent arj the Supervisor, Nuclear
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Operations.
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6.4.3 The Manager, GTL, shall forward the PIT and l
i attached documentation to the Superintendent for evaluation, irrespective of whether the ites is j
i recommended as reportable.
6.4.4.
If Step 6.4.1 cannot be completed within 30 calendar days from the time the Originator initiates the PIT form (Step 6.1), the Manager, GTL, shall notify the Superintendent of the preliminary evaluation of reportability by the end of tha 30-day schedule. The Superintendent may assume responsibility for the.reportability evaluation at this point er request NUSCO to continue with the final evaluation. Step 6.4.4 shall be repeated until a final recommendation is submitted to the Superintendent in Step 6.4.3.
6.5 section 5.0. Superintendent 6.5.1 The Superintendent shall review the PIT and associated documentation and make a final determination as to whether the item is reportable NED 2.25 Rev. 1 Page 8 of 10 i
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i and indicate the sections of 10CTR50.7 10CTR50.73 or 10CTR50.9(b) that apply.
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6.5.2 The Superintendent shall make any necessary notifiestten to tt.e NRC in conf ormance with:
10CTR50.72 (Verhal report within I hour)'
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10CTR$0.73 (Vritten report within 30 days)
G 10CTR10.9(b) (Verbal report vithin 2 days, followed by written report i
vithin 2 week.s) using applicable station procedures. All 50.73 reports are signe: by the Station Superintendent g
and 50.9(b) reports are prepared for corporate
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I officer signature under NEO Procedure 4.01.
I 6.5.3 The Superintendent should indicate any JCOs. LT.Rs or PIRs that are associated with the RET.
6.5.4 The Super,intendent shall return the RET and i
attached documentation to the Manager, GTL.
6.6 Section 6.0. Manarer. CFL 6.6.1 The Manager. CTL. shall ensure that the RET and l
all documentation supporting the determination of i
reportability/conreportability of a potential A
i significant implication or reportable ites are 4.11 l
signed, dated, and filed per NEO 2.13.
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6.6.2 The Manager. GTL, shall inform the Originator and i
the Originator's Supervisor of the final deter-sination.
6.6.3 The Manager, CTL, upon receipt of a JCO,'shall A
forward the JC0 to the Superintendent and the C
NRS/SNRB for review.
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6.7 Substantial Safety Batard Reporting A potential significa.nt. implication or reportable item may be considered for an SSH evaluation and reporting under the provisions of 10CTR21 and NIO 2.01 only after consideration for reporting in accordance with 20CTR50.72, 20CTR50.73, and 10CTR50.9(b).
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NEO 2.25 Rev. 1 l
Page 9 of 10 l
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7.0 T 0'Jls Tirure No.
Firure T!:le 7.1 Flovchart
- dentificat2on. Evaluation, anc Report:ng of Potential Reportable Items."
7.2 Reportability Evaluatien Form.
[g g B.O AMACF.."ES75 S.A Major Changes froo Previous Revision (Rev. O to Rev. 1) 6 NEO 2.25 Rev. 1 Page 10 of 10
i FIGLE.E 7.1 i
i
- CE.C*FICATION. EVALUATICN. AND EEEOETIN3 i
CF POTENTIAL REPORTABLE ITEMS I
i
@IGINATM Initiates EET.
SPERVIS"E kviews. chects, one valicates U.
I 7
M'1.
N
- t. gs in U eno eastgas Icg ruce*.
Csr.tects Lee tetenonnt.
INTDCENT I
SPERIN"DOENT Ciscesttl es oea MECD station % :eores.
evolustion nesos37 yes-MANAGER, GFL Designates responsicility for tecnnical evaluations.
TECH 4ICA2 KANAGERS/SPERVISORS I
Peefore evolustions.
A Mete reconsenestions.
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If net concaste in l20 cars, orevloe YES preltmana y recommenession to Gn..
A prevsee senerets J"Ds if netoec.
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v MANAGER GFL m ytene U
- W INID M If not conclete in 30 cays, notifies I
h oerintenoont of cre11minary c
MECD te continus recosmoneetton, tecnnical If recorthele. arevloes early notification evolustaan?
g gg 1"f,'Y to Loerintencent ene L osevisor, l
4 g
Macass-QWatsons.
Provides all final U recoernrnestions.
Final k commenestion o S.FERINTDOENT
% views U and deiermines eeoortac1111ty.
If recortanle, notifles MC.
v MANAGER GFL Files & ene essociates oocumentation.
Inforse 0 aginator ano nts Loervisor of A
final oetermination. Pmytoes J:Os to
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boerte.tencent and MB/9Hl.
s Y
ffE/9fB i
kviews J:Os.
kv. 1 i
DED 2.25 Page 7.1-1 of 1 l
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TICURE 7.2 NORTEEAST UTIL!!!ES RIPORTABILITY EVALUATION FORh
'l Unit:
RET Log 4 Subjects 4
1.0 CRICINATCR 1.1 Describe potential significant implication or reportable i
item (attach additional documentation if required).
l 1.2 Describe potential safety implications.
i 1.3 Identify applicable sections of 20CTR50.72,' 10CTR50.73, l
or 10CTR50.9(b).
1.4 Torward RET to Immediate Supervisor.
3 9
Originator Date SUPERVISOR OT ORIGINATOR Date Received Suggested Maximus Processing Time: 1 working day 1.5 confirm that the originator has satisfactorily completed i
Steps 1.1 through 1.3.
1.6 Torvard to Manager, GTL.
1 i
Supervisor Date NEO 2.25 Rev. 1 i
Page 7.2-1 of 4
.0 MANAGER. CTL Date Received Sc.ggested Max 2mus Processing Time: three working days.
2.1 C=ntact Superantendent to determine disposition of potential significant implication or repottable item (select options).
NUSCO reportability evaluation requested.
JC0 evaluation requested.
Potential significant implicatien or reportable item vill be dispositioned in accordance with applicable station procedures.
Ferson contacted:
Date:
2.2 Designated Evaluator (s)
Manager. GFL Date 3.0 TECENICAL SUPERVISOR /
Date Received MANAGER Suggested Maximum Processing Time No more than 20 calendar days from Step 1.4, to complete all' technical evaluations.
Supervssor/ Manager Date Not Reportable JC0 Required Reportable per NIO 2.25 Rev. 1 Page 7.2-2 of 4
1 s.
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Supervisorinanager Cate r
Not Reportable JC0 Required Reportable per i
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Supervisor /nanager Date Not Reportable JC0 Required Reportable per f
4.0 MANAGER. GENERATION Date Received FACIL:!*ES LICENSING l
Suggested Maximus Processing Times three working days.
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4.1 Not Reportable Reportable per i
4.2 If yes, provide early notification to the Superintendent and the Supervisor, Nuclear Operations by telephone.
f Date i
Manager GFL Date 1
P r
1 6
5.0 SUPERINTENDENT Date Received 5.1 Iten determined to be reportable?
1 No
- Yes, Section(s) i i
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NEO 2.25 Rev. 1 1
Page 7.2-3 of 4 i
5.2 If ites is deterraned to be reportable, follow applicable station procecures for FRC reporting requirements.
10CTR30.72 items are verbally reported vitnin 1 hourt 10CTR50.73 items are suceitted unoer Statson Superantendent signature vathin 30 days; and 10CTK50.9(b) items are ver-bally reported within 2 days then suositted under corporate officer signature within 2 weeks.
Applicaele PIR/LER Number Unat Superantencent Date l
6.0 MANAGER. CTL-Date Received 6.1 If JC0 is performed, send REF/JC0 to NRB/SNRB. f or review.
Date Sent:
6.2 Ensure documentation completed, logged, filed and i
distributed.
Ey Date:
l et Originator Originator's Supervisor Nuclear Records j
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NE0 2.25 Rev. 1 Page 7.2 4 of.4 i
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l ATTACEMENT 8.A i
KAJOR CP.ANCES TPOM PP.E'.*IOUS REVISION I
i (REV. O UFDA!IO TO REV. 1) l l
Corresp.
No.Date Sender /
Change Receiver Corresp.
Section No.
Description of Number PAR No.
Subject Affected Chante j
1 E.J. Mroczka Timeliness 3.1.4, 5.3.1, Provide report-Meso of Report-5.3.3, 6.0, ability evalua-NED-88-G-170, ability 6.3.1, 6.3.2.
tions to super-March 4, 1988 Evaluations 6.4.1, 6.4.4, intender.t within PAR 88-119-1 Fig. 7.1 & 7.2 30 days of ini-tiation of an evaluation.
When final evaluation can-not be completed within 30 days, provide a pre-liminary evalu-ation and estimate of final comple-
)
tion within 30 days.
2 E.J. Mroczka Memo significant Title. 3.1.4, Reference NEO-88-G-081 Implications 4.2, 4.3, 4.4, 10CrK50.9, February 1, 4.7, 5.2.1, and implement 1988 5.2.2, 6.1.1, instructions 10CTR50.9(b) 6.1.1.1, for performing PAR 89-119-1 6.1.1.2, 6.1.2, "significant 6.2.4, 6.2.5, implications" 6.3.1, 6.5.1, evaluations.
6.5.2, 6.6.1, I
6.7 and Fig. 7.2 3
FAR 88-119-1 JCOs 4.1, 4.6, Describe the l
MP3 EE0 Audit 5.1.5, 5.2.3, JC0, and detail-5.2.4, 5.3.4, instructions 5.3.5, 5.4, for performing 6.1, 6.2.6 JCos.
6.3.3, 6.5.3.
6.6.3. Tig. 7.1 NEO 2.25
- a.
<a 0 1 of 1 l
Docket No. 50-245 B14692 - Exhibit 7 Millstone Nuclear Power Station, Unit No. 1 December 1993
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MNPS-1 UTSAR CHAPTEF 6 TABLE OF CONTENTS
_e 7 -.. e.,
n.,.. e rA--
u 6.0 ENGINEERED SAFETY TEATURES 6.1-1 6.1 ENGINEERED SAFETY FEATURE MATERIALS.............
6.1-1 6.1.1 Metallic Materials 6.1 '.
6.1.2 Organze Materials 6.1-4 6.2 CONTAINMENT SYSTEMS 6.2-1 i
6.2.1 Primary Containment Functional Cesign 6.2-1 6.2.2 containment Heat Removal..........................
6.2-25 6.2.3 Secondary Containment Functional Oesign
. 6,2-2B 6.2.4 Containment Isolatien System 6.2-37 6.2.5 Containment Capability with Respect to Metal-
- ' ate r Reactions
................................. 6.2-39 n
References 6.2-45 6.3 EMERGENCY CORE COOLING SYSTEMS (ECCS) 6.3-1 6.3.1 Design Bases 6.3-1 6.3.2 System Design..................................... 6.3-5 6.3.3 Performance Evaluation 6.3-17 6.3.4 Tests and Inspections 6.3-22 6.3.5 Instrumentation Requirements 6.3-24 References 6.3-26 6.4 HABITABILITY SYSTEMS 6.4-1 6.4.1 Design Bases 6.4-1 6.4.2
System Design
6.4-1 6.4.3 System Operational Procedures 6.4-4 6.4.4 Design Evaluations 6.4-4 i
6.4.5 Testing and Inspection 6.4-4 6.4.6 Instrumentation Requirement 6.4-5 1
References 6.4-6 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5-1 l
6.5.1 Engineered Safety reature (EST) rilter Systems 6.5-1 6.5.2 Containment Spray System 6.5-6 i
Page 6-i March 1988 i
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t MNPS-1 UTSAR CHAPTER 6 i
TABLE OF CONTENTS l
[ CONTINUED)
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.e r -.. -..
7,r pgu.
u....
4--
t.
6.5.3 rissien Product Centrol Structures and Systems 6.5-s References.......................................: 6.5-7 1
1 6.6 IN-SERVICE INSPECTION OT CLASS 2 AND 3 l
CCMPCNENTS 6.6.............................,
i i
l References
- 6. 6-2.
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6.7 MAIN STEAM LINE ISOLATION VALVE LEAKAGE CONTROL l
SYSTEM 6.7-1 1
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t Page 6-ii March 1988 i
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_...____ _._ __ _. _. _,..._._ _.___s
=
I MNPS-1 UTSAR i
CHAPTER 6 LIST CF TABLES l
TABLE NUMBER TITLE 6.1-1 Furnace-Sensit;:ed Stainless Steel in Millstone Unie :
[
6.2-1 Principal Oesign Parameters of the Primary Containment 6.2-2 Flow Rates for Long Term containment Response 6.2-3 Ma]ct Penetration Classification 6.2-4 Principal Penetrations of Primary Containment and l
Associated Isolation Valves 6.2-5 Isolation Signal Codes 6.3-1 Summary of Provisions for Emergency Core Cooling l
6.3-2 Emergency Core Cooling System Summary l
4 i
6.3-3 Summary of Core Spray System Component Performance and Design Parameters 6.3-4 Summary of Low Pressure Coolant Injection Component Performance and Design Parameters 6.3-5 Summary of reedwater Coolant Injection System Component 3
Performance and Design Parameters 6.3-6 Summary of Automatic Pressure Relief (APR)
System Performance and Design Parameters
~
6.3-7 Summary of Isolation Condenser System Performance.and Design Parameters 6.3-8 Summary of Results of LOCA Analysis 6.3-9 MAPLHGR Versus Average Planar Exposure 6.3-10 ECCS Lead Time Sequence. Gas Turbine (PWCI and Half.of.
ECCS Loads)
Diesel Generator (Normal ECCS) Loading l
Sequence 6.3-11 Significant Input variables Used in Loss-of-Coolant Accident Analysis j
Page 6-iii March 1988
'l
MNPS-1 UTSAR CHAPTER 5 LIST CT TABLES tCCNT!NUED)
TABLE NUMBER TITLE 6.4-1 Millstene Unit One, Integrated 30-Day Whole Body Ceses, REM 6.4-2 MP-1 OBA Assumptions 6.5-1 Maximum Heat Loads On Filter Media for Vaticus Containment Leak Rates Page 6-iv March 1988
MNPS-1 UTSAR CHAPTER 6 LIST or TIGURES FIGURE NUMBER TITLE 6.2-1 Pressure Suppression Containment System t
6.2-2 Bodega Bay Tests - Vessel Pressure and Drywell Pressure 6.2-3 Pressure Response - Calculations & Measurements 6.2-4 Comparison of Calculated and Measured Peak Drywell Pressure 6.2-5 Long Term Centainment Pressure Response for Millstone Unit 6.2-6 Recirculation Line Break - Illustration 6.2-7 Long Term Containment Temperature
Response
for Millstone Unit 6.2-8 Millstone Containment capability 6.2-9 LPCI/ Containment Cooling Schematic Diagram 6.2-10 Center Section - Hot Fluid Piping Penetration Assembly j
i 6.2-11 Center Section - Cold Fluid Piping Penetration Assembly 6.2-12 Main Steam Isolation valve Section 6.2-13 Main Steam Isolation valve - Control Diagram 6.2-14 Main Steam Line (Typical of 4) j 6.2-15 reedwater Line (Typical of 2) 6.3-1 Core Spray System P&ID 6.3-2 Low Pressure Coolant Injection System / Emergency Service
{
j Water 1
1 6.3-3 Condensate and reedwater Systems P&ID l
6.3-4 SRV Arrangement 6.3-5 Isolation Condenser P&ID Page 6-v October 1988 i
w
MNPS-1 UTSAR CHAPTER 6 LIST OF FIGURES (CONTINUED)
FIGURE NUMBER TITLE 6.3-6 Millstone Core Spray Pump Characteristics Curves 6.3-7 Millstone Low Pressure Coolant Injection Pump Characteristic Curves 6.3-8 Peak Cladding Temperature Versus Break Area 6.3-9 Typical Power Generation Following a Design Basis Loss of Coolant Accident 6.3-10 variation with Break Area of Time for Which Hot Node Remains Uncovered 6,4-1 Control Room - General Arrangement and Details 6.5-1 HVAC Flow Diagram - Reactor Building Radioactive Waste Building a
l Page 6-vi October 1988
MNPS-1 UFSAR 6.3 Emereency Core Cooline Systems (ECCS) 6.3.1 Desien Bases This section describes the functicnal requirements and performance characteristics of the engineered safeguards that have been previded in additien to those safety. features included in the design of the reactor, reactor coolant system, unit containment
- system, reactor control systems and other instrumentatien et process systems described elsewhere in this report.
These safeguard features, with the exception of the feedwater equipment, serve no function that is necessary for normal station operatien.
They are included in the plant for the sole purpose of reducing the consequences of postulated accidents.
The emergency core cooling system is automatically placed in cperation whenever a less-of-coolant condition is detected.
The systems contained in the emergency core cooling system are the core spray (CS), low pressure core injection (LPCI), feedwater coolant injection (rWCI), automatic pressure relief (APR:
and the isolatien condenser (IC) system.
j Core Soray System (CS)
The core spray system is a low pressure system designed to coel the reactor core by direct impingement of high-density spray following depressurization.
The core spray system supplies cooling water to the core to prevent fuel clad melting and excessive fuel clad metal-to-water reactions during a LOCA.
The system provides this protection for any pipe break up to and including the largest oossible break, which is a double ended circumferential shear of a
28 in. recirculatien system pump suction line.
This is equivalent to a 4.3 square foot break aiea.
For intermediate-size breaks of 0.01 to 0.20 square feet, i
the automatic pressure relief (APR) system may need to be actuated to rapidly reduce reactor pressure to within the injection pressure range of the Core Spray and Low Pressure Ccolant Injection Systems.
Breaks of this size result in system i
losses which exceed the make-up capacity of the feedwater coolant injection system, but may not be large enough to depressurize the i
reactor before unacceptable fuel temperatures are reached.
The core spray system is a two train system consisting of the following major components-i Page 6.3-1 February 1987
{
1
MNP5-1 UFSAR 1
(1)
Pump sucticr. isolation valve (2)
Co.5e spray pump (3)
Test valve to terus (4)
Admission valves (2)
(5)
Relief valve (6)
Core spray sparger (7)
Suosystem logic Lew Pressure Coolant Injection System (LPCI)
The low pressure coolant injection (LPCI) system provides high-volume emergency makeup to the reactor vessel in the event f
of a loss of coolant accident (LOCA).
The LPCI and emergency j
service water systems together constitute the containment cooling subsystem.
The LPCI system uses electric motor-driven pumps to transfer water from the suppression pool or condensate storage tank to the reactor vessel or the containment spray headers.
The LPCI syst;m is designed to limit fuel damage in the event of a depressurizing, large line break LOCA.
It is capable of adequate cooling far break areas between 0.2 and 4.3 square feet.
The 4.3 j
square icot break area corresponds to the design basis
- LOCA, i
which is a double-ended shear of a 28 in, reactor recirculation system pump suction line.
The LPCI system also provides protection for small and intermediate line breaks, where the feedwater coolant injection system either is not available or cannot supply reactor-vessel makeup requirements indefinitely.
Use of LPCI under these conditions requires the reactor vessel to be depressurized to the LPCI operating
- range, and may require actuation of the safety / relief
- valves, either manually or via the automatic pressure relief system.
The LPCI system is two loop system containing similar arrangements of piping and components.
Each loop LPCI A and B-contains:
(1)
Low pressure coolant injection pumps (2),
(2)
A low pressure coolant injection heat exchanger, i
(3)
A drywell spray header, and (4)
Necessary piping, valves and instrumentation A single torus spray header is shared by LPCI loops A and B.
1 Teedwater Coolant Injection System (FWCI) i The FWCI system is a high pressure emergency core cooling system that utilizes existing feedwater and condensate components (A and Page 6.3-2 February 1987
MNPS-1 UrSAR B
trains) to supply core cooling.
In the e v e r. t of an intermediate loss of coolant accident, the FWCI system will cause plant depressurization as well as assist in maintaining water level in the vessel.
This reduction in plant pressure will allow the low pressure coolant injection ano core spray systems te function to prevent core damage.
The TWCI system utilizes a logic network that integrates the functions of the following:
(1) reedwater system (2)
Condensate system (3)
Emergency condensate transfer System (4)
Gas turbine generator The FWCI system is comprised of the following major components:
(1)
Emergency condensate transfer pump (2)
Condensate pumps (A and B train)
(3)
Condensate booster pumps (A and B train)
(4) reedwater pumps (A and B train)
(5)
Condenser hotwell (6)
Gas turbine generator Automatic Pressure Relief (APR)
The automatic pressure relief system is designed to provide a back method of rapidly reducing plant pressure during a small break loss-of-coolant accident.
Reactor system depressurization cooling is normally accomplished under these conditions and core by the feedwater coolant injection system.
In the event that the feedwater coolant injection system fails during a loss-of-coolant accident the APR system will open selected steam relief valves (SRV) to rapidly depressurl:e the reactor plant to within the injection pressure range of the core spray and low pressure coolant injection systems.
The APR system consists of:
(1)
Safety / relief valve assemblies (4),
(2)
Pneumatic accumulators (4),
(3)
Discharge pipe T-quenchers (4),
Isolation Condenser The isolation condenser system is an emergency core cooling system (ECCS) which can be actuated automatically or manually.
The system removes residual and decay heat from the reactor vessel in the event that the main condenser is not available, or Page 6.3-3 February 1987
MNPS-1 UFSAR a
high reactor pressure condition exists.
The system employs natural circulation as the driving head through the isolation condenser tuoes.
The shell side of the isolation condenser contains a
water volume which betis off to remove heat transferred from the reactor.
The isolation condenser system
=ust also a:d in reactor vessel depressurization in the event that the feedwater coolant injection system fails.
The isolatica condenser system's normal cperating functions and a ecmplete system ecmpenent descriptien are provided in Chapter 5.
The central principle of coolant system design is to provide core cooling continuity over the entire range of operating conditions and postulated accident conditions.
During normal operation, when normal auxiliary power is available, heat is removed from the core through the steam-turbine-condenser cycle or through the use of the shutdown cooling system.
These are the normal provisions for core cooling.
When the reactor is isolated from the main condenser and the shutdown cooling system, or - when electrical power is unavailable to pump cooling water to the condensers and heat exchangers, and in the absence of f.ny loss of coolant from the primary systems, the core is cooled by relief valve action followed by use of the isolation condenser
- However, other means are needed to provide continuity of core cooling during those postulated accident conditions where it is assumed that mechanical failures occur in the primary system and coolant is partially or completely lost from the reactor vessel and either (1) normal auxiliary power is unavailable to drive the feedwater pumps or (2) the loss of coolant occurs at a rate beyond the capability of the feedwater system.
Under these circumstances, core cooling is accomplished by means of the emergency core cooling system ( ECCS) which consists of two core spray subsystems, the TWCI subsystem, the LPCI/ containment cooling subsystem, the automatic pressure relief subsystem, and the isolation condenser.
The overall ECCS design bases are:
(1)
The emergency core cooling system is designed to prevent fuel cladding melting for any mechanical failure of the primary system up to and including a break area equivalent to the largest primary system pipe.
(2)
The entire spectrum of line breaks, up to and including l
this maximum, is designed to be protected against by at j
least two independent cooling methods which are i'
actuated automatically.
(3)
No reliance is assumed to be placed on external sources of power.
Page 6.3-4 February 1987
I MNPS-1 UrSAR i
(4)
The emergency core cooling system is capable cf fulfilling its performance function under the most adverse of postulated accident conditions.
The emergency core cooling system is designed to protect the reactor core against cladding meltdown across the entire spectrum l
of line break accidents.
Table 6.3-1 summart:es the provisions for emergency cooling of the reactor core under various conditions.
A summary of the emergency core cooling subsystem i
parameters is shown in Table 6.3-2.
The individual system components which comprise the ECCS are described in Section 6.3.2 and the integrated performance is evaluated in Section 6.3.3.
The reliability considerations incorporated into the design of the ECCS are discussed in Section 6.3.2.5.
Section 3.5 describes the effects of missiles on the plant.
The capability of the EC~S to withstand the effects associated i
with postulated pipe ruptures is described in Section 3.6.
A description of the ECCS seismic analysis is provided in Secticn l
3.7.
l The environmental design bases with respect to high temperature steam atmosphere and containment sump water level that might exist in the containment during ECCS operation is discussed in Section 3.11.
6.3.2 System Desien 6.3.2.1 Schematic Pipinc and Instrumentation Diacrams i
The CS,
- LPCI, TWCI,
- APR, and IC piping and instrumentatien diagrams are shown in rigures 6.3-1, 6.3-2, 6.3-3, 6.3-4, and 6.3-5, respectively.
The ECCS interlocks and their function are as follows:
(1)
Core Soray System Interlocks runction APR (ac power)
Prevents auto-blowdown unless the Interlock discharge pressure of at least one core spray or low pressure coolant injection pump is 100 psig or greater.
1 Page 6.3-5 February 1987 I
t MNPS-1 UTSAR CS In:tiatien a.
Starts CS pumps at
-48
- n.
vessel level or 2 psig drywell pressure.
b.
Interlocks test to torus valves i
closed.
c.
Interlocks pump suction valves open.
d.
Interlocks admissien valves open if reactor low pressure permissive is satisfied.
Reactor Low At 350 psig decreasing, permits Pressure the admission valve to open with Permissive an automatic initiation signal present.
j 1
Close Permissive Allows sealed-open valves to be l
closed if the A(B)
CS/LPCI Close Permissive Switch is in the CLOSE PERMISSIVE position.
i (2)
Low Pressure Core Injection System Interlocks Function l
APR (ac power)
Prevents auto-blowdown unless the Interlock discharge pressure of at least one core spray or LPCI pump is 100 l
psig or greater.
i LPCI Pump Minimum Open the appropriate minimum flow.
l Flow bypass to torus when the flow from any LPCI pump is 250 gpm or less.
i LPCI Initiation
- a. Starts LPCI pumps at
-48 in.
vessel level or a high drywell pressure of 2 psig.
- b. Interlocks containment spray and test valves closed.
I l
i Page 6.3-6 December 1987 l
4 MNFS-1 UTSAR c.
Interlocks heat exchanger bypass valves cpen for 1
minute.
LPCI Loop Select:en At -48 in.
vessel level or 2 psi drywell pressure:
a.
Selects loop for injection based on reactor recirculatien pump and jet pump riser dP.
b.
Interlocks injection valves in the selected loop open for 5
minutes after reactor pressure deer. eases to 350 psig.
c.
Interlocks inboard injection valves in the nonselected loop closed for 10 minutes.
- d. Shuts the reactor recirculation pump discharge valve in the selected loop.
t Reactor Low Functions with LPCI loop selection Pressure Permissive logic to interlock LPCI injection valve closed until reactor pressure is 350 psig or less.
Prevents simultaneous opening cf both injection valves in a LPCI loop unless pressure is 350 psig or less.
Containment Spray Allows drywell and torus spray a LPCI Interlock stops to be opened with initiation signal present, if:
- a. Containment spray first key is in manual override,
- b. Vessel water level is at least 2/3 core height or containment spray second key is in manual override.
Page 6.3-7 February 1987 i
i
MNPS-1 UTSAR
- c. Drywell pressure is 5 psig or greater.
Suppression Pool Allows test valves to torus to be Coolin; Interlock opened with a
LPCI initiatien i
signal present if:
a.
Containment spray first key is in manual override, b.
vessel water level is at least 2/3 core height or containment spray 2nd key is in manual override.
Close Permissive Allows sealed-open injection Interlock valves to be closed if the respective CS/LPCI Close Permissive Switch is in the CLOSE PERMISSIVE position.
(3) reedwater Coolant Iniection System Interlocks runction FWCI initiation The TWCI system is initiated automatically by a
reactor pressure vessel low-low water level of -48 in. or a high drywell pressure of 2 psig.
The following component actuation occurs when the above signals are present:
- a. Starts emergency condensate transfer pump.
- b. Starts the preselected A or B string condensate pump.
c.
Starts the preselected A or B string condensate booster pump.
- d. Starts the preselected A or B string reactor feed pump.
e.
Emergency condensate transfer pump discharge valve opens.
Page 6.3-8 February 1987
MNPS-1 UFSAR
- f. Feedwater regulating valve C
train closes and A and B train feedwater regulating valves open.
g.
Condensate supply valves to recombiners shut.
h.
Steam jet air e]ector minimum flow recirculatten valve closes.
- i. Reactor feed pump minimum flow recirculation valve shift to automatic control.
Main Condenser Prevents automatic initiation of Hotwell Level FWCI if level in either hotwell is less than 12 in.
Feed Pump Pressure a.
Prevents feed pumps A
and B from starting unless suction pressure is greater than 335 psig.
t b.
Close feedwater regulating valves at 300 psig decreasing discharge pressure if there is a
loss of normal power concurrent with a
FWCI initiation signal.
c.
Reset the feedwater regulating valves from flow control to level control at 300 psig decreasing discharge pressure.
(4)
Automatic Pressure Felief System Interlocks runction Auto Blowdown APR initiation is prevented unless System A/C an LPCI or CS pump is running with Interlock 100 psig discharge pressure, i
i l
l i
Page 6.3-9 February 1987
i MNPS-1 UTSAR
{
(5)
Isolation Cendenser System Interlocks Function IC initiation Opens cendensate isclation velve and closes the normally open steam vent valves at
-48 in, reacter l
water level or reactor pressure greater than 1085 psig for 15 seconds.
6.3.2.2 Eculpment and Component Descrittiens see Section 3.11 for a discussion of environmental qualification of ECCS equipment and compenents.
6.3.2.2.1 Core Sprav Svstem Core Scrav Fumps j
t The core spray pumps are vertical, single stage, double suction pumps with inline suction and discharge nor:les.
Each pump is driven by a vertical motor through a flanged spacer coupling.
Shaft sealing is accomplished by a mechanical ceal that has been specially designed with a
secondary seal gland cover to allow collection and drainage of any seal leakage including leakage from serious failure of the mechanical seal.
Clean flushing is provided to the seal faces and the internal bearing by a bleed off line taken frem the pump discharge and passing it through a separator.
This clean liquid, under pressure, is first injected into the center of the internal bearing with some passing up into the seal chamber, flushing the seal faces, and returning to the l
suction not:le.
The impeller is. hydraulically
- balanced, both axially and radially.
The axial balance is achieved by the double suction design and having the same diameter wear rings on either side of the impeller.
The deadweight of the rotating element is carried by the thrust bearing in the driver.
The impeller radial balance is achieved by having a double volute construction in the casing.
The driver and pump are connected with a rugged motor stand designed to withstand the design seismic forces.
The complete unit is supported on mounting feet attached to the bottom of the volute casing.
The mechanical seal is a Durametallic type P.T.O. mounted on seal sleeve and held in position with shaft nut.
The shaft nut is designed to act as a slinger in the secondary seal gland cover.
The seal is easily and quickly removed for maintenance frem the stuffing box with the aid of the sleeve mounted seal design.
j i
Page 6.3-10 February 1987 i
F MNPS-1 Ur5AR See Table 6.3-3 for a
summary of performance and desien parameters, and ricure 6.3-6 fer a performance character:stic curve of the core spray cump.
Meter Operated Valves The motor-cperated valves in the cere spray system are conventional glete valves (test linet er gate valves with electric remote centrelled meter-driven operators.
The meter cperators have limit switches to centrol valve travel and terque switches to previde c.:ntrol of seating pressure.
Relief Valves To prevent the possibility of overpressurizing the discharge piping due to leakage from the reactor past the motor operated adm:ssion valve, pressure relief valves are provided.
Relief valves are sized to acec=medate anticipated leakage and piping instrumentation alarms on high pressure.
Pipinc See Table 6.3-3 for a summary of piping design parameters.
Spray Header,s, Eee Table 6.3-3 for a summary of design parameters regarding the spray headers.
6.3.2.2.2 Low Pressure Coolant Injection (LPCI) System LPC: Pumes The LPCI/ containment cooling pump is a vertical, single stage, double suction pump with in-line suction and discharge no::les.
The pump is driven by a vertical motor through a flanged spacer coupling.
Shaft sealing is accomplished by a mechanical seal that has been specially designed with a secondary seal gland cover to allow collection and drainage of any seal leakage including leakage from serious failure of the mechanical seal.
Clean flushing is provided to the sral faces and the internal bearing by a bleed off line taken from the pump discharge and passing it through a
separator.
This clean
- liquid, under
- pressure, is first injected into the center of the internal bearing with some passing up into the seal chamber flushing the seal faces and finally returns to the suction nozzle.
The pump shaft is stainless steel with a bronze impeller keyed en a
tapered hub.
The impeller is hydraulically balanced both axially and radially.
The axial balance is achieved by the double suction design, having the same clameter wear rings en Page 6.3-11 February 1987
MNPS-1 UFSAR either side of the impeller.
The deadweight cf the rotatinc element is carried by the thrust bearing in the driver.
The impeller radial balance is acnieved by having a deucle volute construction in the cast steel casing.
The driver and pump are connected by a
rugged motor stand.
The complete unit is supported en mounting feet attached to the bettem of the volute-Casing.
The pump unit was built to ASME Section III Class C.
The pumps designed to function satisfactorily for any pcst-LOCA fluid are condition.
Pump performance requirements are chosen so that with considerations of elevation
- head, reactor pressure, frictional losses and potential leakage paths within the reacter, total system function can be obtained with three of the four system pumps operating.
See Table 6.3-4 for a
summary cf performance and design parameters, and Figure 6.3-7 for a performance characteristic curve for a single LPCI pump.
Motor-Ocerator Valves The motor-operated valves in the LPCI/ containment cooling system are conventional globe stop-check or gate valves with electric motor-driven operators.
The motor operators are operated via limit switches and torque switches to provide autematic remote control as needed.
Except for the LPCI injection valves and containment spray valves, all valve stroke rates are standard speeds.
All system valves are designed in compliance with the pressure and temperature ratings of the line to which they are attached.
In the area of a change of pipe pressure rating, the valve is designed to the higher requirements.
Valve body materials are consistent with the materials of the piping to which they are attached.
Emercency Service Water Pumps Two emergency service water (ESW) pumps supply cooling water to each of the two containment cooling heat exchangers.
At rated conditions they supply cooling water at a pressure higher than the corresponding process water, preventing outleakage from the primary system.
See Table 6.3-4 for a
summary of performance and design parameters.
Page 6.3-12 February 1967
'I o
l l
l
(
MNPS-1 UTSAR
+
Heat Exchancer LPCI Heat Exchangers The LPCI heat exchangers are
.ertically mounted, single-pass, t
shell-and-tuce heat exchangers, each rated for a heat transfer i
capacity of 40 million Stu/hr.
LPCI system coolant flows through the snell side and gives up heat to Emergency service Water l
counterflow en the tube side.
1 Emergency Service Water pressure is maintained at least 15 psi greater than LPCI System pressure during operation, and the emergency service water heat exchanger outlet valves are interlocked shut when an associated emergency service water pumps i
are stopped.
These measures are taken to prevent an unplanned release of potentially contaminated water via the Emergency Service Water System if a
heat exchanger tube leak should develop.
See Table 6.3-4 for a summary of performance.
6.3.2.2.3 reed Water Coolant Injection (rWCI) System See Table 6.3-5 for a
summary of performance and design parameters en equipment utili:ed in the FWCI system.
6.3.2.2.4 Autematic Pressure Relief (APR) System Safety Relief Valves The Target Rock Model 7567F pilot-operated safety / relief valve l
consists of two principle assemblies:
a pilot stage assembly and the main stage assembly.
These two assemblies are directly I
coupled to provide a unitized, dual function safety / relief valve.
l The pilot stage assembly is the pressure-sensing and centrol I
- element, and the main stage assembly is a system fluid-actuated follower valve which provides the pressure relief function.
Self-actuation of the pilot assembly at set pressure vents the main piston chamber, permitting the system pressure to fully open the main assembly which results in system depressurization at l
full rated flow.
(
The pilot assembly of the Target Rock safety / relief valve consists of two relatively
- small, low flow pressure-sensing i
elements.
The spring-loaded pilot disc senses the set pressure, l
and the pressure-loaded stabilizer disc senses the reseat 1
l pressure.
Spring force (preload force) is applied to the pilot i
disc by means of the pilot rod.
Thus, the adjustment of the i
spring preload force will determine the set pressure of the i
valve.
1 The main assembly of the Target Rock safety / relief valve is basically a reverse (pressure) seated, system fluid-actuated Page 6.3-13 February 1987 l
l 1
t I
MNPS-1 UFSAR l
angle globe valve.
Actuation cf the main assembly permits discharge of fluids frcm the protected system at the valve's rated flow capacity and provides the system pressure relief i
function of the valve.
The major ccmponents of the main stage 3
are the valve tedy, disc /pisten assembly and preload spring.
l See Table 6.3-6 for a
summary of performance and design parameters of the APR System.
6.3.2.2.5 Isolation Condenser f!C) System j
!solatien Condenser The isolation cendenser is a large shell and U-tube type heat exchanger with steam or condensate on the tube side and water en the shell side.
The shell side is three-fourths filled with water and is vented to the atmosphere through the reactor building wall.
The Inconel tubes make up two bundles of U-tubes, one en each side of the isolation condenser.
The isolatien condenser is sized to accommodate the fully decay heat load five minutes after a scram, assuming the maximum decay heat load.
The shell water volume is based en approximately 30 minutes of condensing following reactor isolation without additional makeup water.
The tube bundles in the isolation condenser are elevated about 25 feet above the normal reactor vessel water level.
This elevation provides sufficient driving force (head) to operate the system en gravity alone.
The normal supply of water to the isolation condenser is from the fire main system and a
backup supply is provided by the condensate transfer system.
The normal level in the isolation condenser is about three-fourths full of water.
See Table 6.3-7 for a summary of the design and performance parameters for the Isolation Condenser System.
6.3.2.3 Applicable Codes and Classifications The applicable codes and -classifications of the ECCS are specified in Section 3.2 Equipment, piping and support structures of CS, LPCI, TWCI, APR and the IC have been designed in accordance with Class I seismic j
criteria.
This is discussed in Sections 3.2 and 3.10.
IEEE codes applicable to the controls and power supplies are specified in Section 7.3.4.1.2.
1 i
i l
Page 6.3-14 February 1987
)
l
MNPS-1 UFSAR 6.3.2.4 Material Specificatiens and Ccmpatibility_
discussed in Section 5.2.3.
LPCI materials are discussed in ECCS materials are discussed in Section 6.1.
APR materials are Section 5.2.3.
The in-vessel portions of the CS system are discussed :n Section 4.5.2.
6.3.2.5 System Reliability
[
Analysis shows that no single failure prevents either the starting of the ECCS when required, or the delivery of ecolant te the reactor vessel.
The ECCS design criterion of no clad melting is satisfied across the entire spectrum of possible liquid er steam line break sizes by at least two separate and independent systems and by two different modes of core cooling even in the event of the loss of normal station power with only one on-site power source available.
In addition, redundancy in the system exists since only one of the two 100 percent capacity core spray systems or only three of the four available LPCI pumps are needed for large breaks; protection for small breaks is provided by the TWCI or auto blowdown followed by operation of either the core spray or LPCI systems.
6.3.2.6 Protection Provisions Protection provisions have been included in the design of the ECCS.
Protection has been provided against missiles, pipe whip,
- flooding, thermal stresses, and the effects of LOCA and seismic events.
Protection of the ECCS piping and components against internally and externally generated missiles is discussed in Section 3.5.
A discussicn is provided in Sections 3.5 and 6.3 describing the effects of postulated pipe ruptures and pipe whip with respect to ECCS.
Flooding protection is discussed in Section 3.4.
The methods used to provide assurance that thermal stresses do not cause damage to the ECCS are described in Section 3.9.
The component supports which protect against damage from movements and seismic effects are described in Section 3.7.3.
6.3.2.7 Provisions for Performance Testing The core spray system has been designed to permit testing the performance of its various components.
The core spray system is a
two train system.
A test line capable of full system flow is connected from a point near the outside isolation valve back to the torus.
Flow can be diverted into this line to test the Page 6.3-15 rebruary 1987
y j
MNPS-1 UFSAR 1
operability control system and pumps during reactor operatten.
{
Water from the condensate storage tank is used for system flusning and suppression pool water is used for periodic testing of the system.
The LPCI system has been designed to permit testing of its components while the unit is operating or pressurized at hot standby.
The LPCI pumps can be started and full flow established through the bypass line back to the pressure suppression pool to determine availability of pumps and control circuits.
The motor-operated valves can be exercised and their operability demonstrated.
Leak-tightness of the system can be demonstrated.
When the unit is shutdown and depressurized, flow rate measurements can be made with water pumped into the reactor vessel.
Also, relief valves on the lower pressure lines can be removed and tested for set point.
The FWCI system is a two train system.
The FWCI system is designed so that those components of the system not normally in service can be tested on a
periodic basis to demonstrate availability of the system.
The condensate and feedwater pumps and systems are in use during normal plant operation and are therefore continuously monitored for malfunctions and deterioration.
Each APR system relief valve.can be opened manually to verify operability.
Testing of the APR system is done at frequency that assures availability of the system.
The motor-operated valves of the isolation condenser system are tested on a routine basis for operability.
Simulated automatic actuation and functional system testing are performed during refueling outages.
6.3.2.8 Manual Actions The ECCS is actuated automatically and requires no operator action during the first 10 minutes following the accident.
i During the long-term cooling period (after 10 minutes), the operator takes action, as specified in Section 6.2.2.1, to place the containment cooling system into operation.
Placing the containment cooling system into operation is the only manual action that the operator needs to accomplish during the course of the LOCA.
The operator has multiple instrumentation available in the main control room to assist him in assessing the post-LOCA conditians.
This instrumentation indicates reactor vessel pressure and water
- level, and primary containment
- pressure, temperature, and radiation levels, as well as indicating the operation of the Page 6.3-16 February 1987
s
.<.-v.-.
me MNPS-1 UTSAR t
ECCS.
ECCS flow indicatzen :s the primary parameter available to proper operatien of the system.
Other indications, such assess as position of valves, status of circuit breakers, and essential power bus voltage, are also available to assist the operator in determining system operating status.
The instrumentatien and controls for the ECCS are discussed in Section 7.3.
Monitor:nc instrumentation available to the operator is discussed in more detail in Section 7.7.
Consideration has been given to the unlikely possibility that manual valves in the ECCS might be left in the wrong position and remain undetected when an accident occurs.
Many of the manual valves in the ECCS are vent, drain, or test connection valves, which are normally closed and capped.
Administration centrols, such as prestart-up valve lineup
- checks, are sufficient to reasonably ensure that such valves will not degrade ECCS performance.
Certain other valves are physically locked in their
~
normal position.
Access to the keys to
.he locks is controlled administratively.
In other
- cases, two isolation valves are provided in series to minimize the possibility of inter-or l
intra-system leakage.
Position indication of manual valves that are in the main flowpaths of the ECCS, and that are inaccessible during normal plant operation, as provided in the main control room.
Proper administrative controls and/or surveillance testing are relied upon to ensure the position of the remaining valvec.
6.3.3 Performance Evaluation The performance of the emergency core cooling system (ECCS) is determined through application of the 10 CTR 50, Appendix K evaluation
- models, and demonstrated by conformance to the acceptance criteria of 10 CTR 50.46 (Reference 6.3-1).
The ECCS performance is evaluated for the entire spectrum cf break sizes for postulated LOCAs.
The accidents, as listed in Chapter 15, for which ECCS operation is required are:
(1)
Section 15.7.4 Loss of Coolant Accident (LOCA)
(2)
Section 15.7.6 Loss of feedwater flow Chapter 15 provides a
description of the radiological consequences of the events listed above.
i Page 6.3-17 February 1987 i
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MNPS-1 UFSAR 1
I 6.3.3.1 ECCS Bases for Technical Specificatiens i
The maximum average planar linear heat generation (MAPLHGE) i calculated in this performance analysis provide the basis ratios for technical specifications designed to ensure conformance with the acceptance criteria of 10 CTR 50.46.
Minimum ECCS functienal requirement are specified in Sections 6.3.3.4 and 6.3.3.5.
Testing requirements are discussed in Section 6.3.4.
6.3.3.2 Acceotance criteria for ECCS Performance The applicable acceptance criteria, extracted from 10 CTR 50.46, are listed below.
For each criterion, applicable parts cf Section 6.3.3 (where conformance is demonstrated) are indicated.
detailed description of the methods used to show compliance is Acontained in Reference 6.3-1.
(1)
Criterion 1,
Peak Cladding Temperature "The calculated maximum fuel element cladding temperature shall not exceed 2200*F."
Conformance to Criterion 1 is shown in Sections 6.3.3.7.3, 6.3.3.7 4,
6.3.3.7.5, 6.3.3.7.6, and specifically in Table 6.3-8.
(2)
Criterion 2,
Maximum Cladding oxidation "The calculated total local oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation."
Conformance to Criterion 2 is shewn in Tables 6.3-8 and 6.3-9.
(3)
Criterion 3,
Maximum Hydrogen Generation "The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react."
Conformance to Criterion 3 is shown in Table 6.3-8.
(4)
Criterion 4, Coolab" Geometry
" Calculated changes in core geometry shall be such that the core remains amenable to cooling."
As described in Reference 6.3-1, conformance to Criterion 4
is demonstrated by i
conformance to Criteria 1 and 2.
1 (5)
Criterion 5, Long-Term Cooling "After any calculated successful initial operation of the
- ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed from the extended period of time required by the j
long-lived radioactivity remaining in the core."
Conformance to criterion 5 is demonstrated generically Page 6.3-1B February 1987 i
MNPS-1 UFSAR for General Electric SWRs in Reference 6.3-3.
Briefly summariced, the core remains covered to at least the 3et pump suction elevatten, and the uncovered region is cooled by spray cooling and/or by steam generated in the covered part cf the core.
6.3.3.3 Sincle railure Considerations The functional consequences of single
- failures, including operator errors that might cause any manually controlled.
electrically cperated valve in the ECOS to move to a position that could adversely affect the ECCS, and the potential. for submergence of valve motors in the ECCS, are discussed in Section 6.3.2.6.
The most severe single failures are identified in-Reference 6.3-2.
Therefore, only these single failures.are considered in the ECCS performance analyses.
6.3.3.4 System Performance Durinc the Accident In general, the system response to an accident can be described as the following:
(1)
An initiation signal is received (2)
A small lag time (to open all valves and run the pumps up to rated speed) occure.
(3)
The ECCS flow enters the reactor vessel.
Key ECCS actuation set points and time delays are provided in the Technical Specifications.
The maximization of the delay from the receipt of signal until the ECCS pumps have reached rated speed is limited by the physical constraints on accelerating the standby diesel generator (SDG),
Standby Gas Turbine Generator, and the ECCS pumps.
The delay time due to valve motion in the high-pressure system provides a suitable conservative allowance for valves available for this application.
An interlock is provided on the low pressure ECCS that prevents opening the injection valves unless the reactor pressure is below a preset l
value.
Head-flow curves for core spray and low pressure coolant injection pumps are shown.
in Figures 6.3-6 and 6.3-7, respectively.
Piping and instrumentation diagrams for the ECCS injection network are identified in Section 6.3.2.
Functional control diagrams (TCDs) for the ECCS are provided in Section 7.3.
3 The operational sequence of the ECCS for the design basis LOCA is shown in Table 6.3-10.
Page 6.3-19 February 1987 5
4 4
MNPS-1 UFSAR i
Operator action is not required. except as a monitoring function, during the short-term cooling period follcwing the LOCA.
During the long-term ecoling
- period, the operator takes acticn as specified in Section 6.2 to place the centainment cooling system i
into operatien.
The reedwater Ceolant Injecticn system (FWCI) utilices equipment l
associated with the feedwater system and the condensate system.
Upon loss of effsite power, power supply to the buses feeding this equipment is transferred to the standby diesel generater i
and/cr the standby gas turbine generator.
An emergency condensate transfer pump is automatically placed into service to partially supplement condensate inventories beiag removed frcm i
the condenser.
The autceatic pressure relief system's safety relief valves I
function to support ECCS in the pressure relief mode and function to relieve over pressure in the safety mcde.
The safety / relief j
valve's palet stage senses system pressure for self actuation and provides the control element for actuation from reactor protection circuitry.
The method for self-actuation is the t
safety mcde and the method of centrol element actuatien in the ECCS or pressure relief mode are
- parallel, non-interfering methods of actuating the sane valve.
The isolatien condenser is employed in the same manner in its ECCS mode as it is employed in normal operations.
The system utilices station batteries when being placed in operation.
Group IV isolation signals cause isolation independent of mode cf operation.
6.3.3.5 Limits on ECCS System Parameters Refer to Sections A.6.3.3.6 through A.6.3.3.7.2 of Appendix A of Reference 6.3-2.
Compliance with Regulatory Guide 1.47 is identified in Section 1.8.
6.3.3.7 ECCS Analyses for LOCA 6.3.3.7.1 LOCA Analysis Procedures and Incut Variables Refer to Section A.6.3.3.7.1 of Appendix A,
of Reference 6.3-3.
The significant input variables used by the LOCA codes are given in Table 6.3-11 and on rigure 6.3-9.
4 6.3.3.7.2 Accident Description Reference to a detailed description of the LOCA calculation is i
provided in Section 3.2.5.2 of Reference 6.3-3.
Page 6.3-20 February 1987 i
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4 MNPS-1 UFSAR 6.3.3.7.3 Break Scectrum Calculatiens A
complete spectrum of pestulated break sizes and locations is j
considered in tne evaluation cf ECCS performance.
1 j
A summary of the results of the break spectrum calculations is shown in tabular form in Table 6.3-8 and graphically on rigure 6.3-8.
Conformance to the acceptance criteria (peak cladding temperature (2200'r, peak local exidation 117%, and core-wide metal-water reaction i li) is demonstrated.
A summary cf provisions for emergency core cooling is given in Tables 6.3-1 and 6.3-2 for various breaks, power supply and ECCS availability assumptions.
Details of calculations for specific breaks are included in the subsequent paragraphs.
6.3.3.7.4 tarce Fecirculation Line Break Calculations The characteristics that determine which is the most limiting large break are:
(1)
The calculated time for reficoding the het nede (2)
The calculated time for uncovering the hot node (3)
The calculated time of boiling transition.
The calculated time of boiling transition increases with decreasing break size, since the time of uncovering of the jet pump suction
- inlet, which leads to boiling transition, is r
determined primarily by the break size.
The calculated time for uncovering the hot node also generally increases with decreasing break
- size, since it is determined primarily by the reactor coolant inventory lost during the blowdown.
The hot node reflooding time is determined by a number of interacting phenomena, such as depressurization
- rate, countercurrent flow limiting, and a combination.of available ECCS.
The period between the uncovering of the hot node and its reflooding is the period when the hot node has the lowest heat transfer.
Hence, the break that results in the longest period during which' the hot node remains uncovered results in the highest temperature.
If two breaks have similar times during which the hot node remains uncovered, then the larger of the two breaks will be
- limiting, as it would 'have an earlier boiling transition time (i.e.,
the larger break would have a more severe result from a blowdown heat transfer analysis).
Page 6.3-21 February 1987
MNPS-1 UTSAR rigure 6,3-10 shows the variation with break size cf the calculated time the het node remains uncovered.
Based on these calculations, the design basis accident (DBA) was determined to be the break that results in the highest ca{culatedpeak cladding temperature in the 1.0 ft' to 4.3 ft region (the largest pessible area of a ree:rculation system line break is 4.3 ft').
The maximum average planar linear heat generation rate (MAPLHGR),
maximum local exidation, and peak cladding temperature as functions of exposure (from the analysis of-the DBA),
are summarized in Table 6.3-9.
l 6.3.3.7.5 small Recirculation Line Break Calculations 2
for small recirculation line breaks of 0.10 ft and belew, a
discharge break coincident with a
gas turbine failure and injection of LPCI into the broken recirculation loop is the most limiting break / failure confirmation.
As summarized in Reference 6.3.1, this results in a peak clad temperature that is less than that calculated for the large break and is bounded by the Design Basis Accident (DBA).
For small recirculation line breaks between 0.10 ft and 1.0 ft the LPCI injection valve failure is the most limiting failure." A low flow MAPLHGR multiplier of 0.95 was derived for small break analysis (see reference 6.3-4).
6.3.4 Tests and Inspections t
6.3.4.1 ECCS Performance Tests preoperational test of the ECCSs was Prior to plant startup, a conducted.
This test assured the proper functioning cf all instrumentation, pumps, heat exchangers and valves and verified i
that the system met its design performance requirements.
The preoperational tests performed on the emergency core cooling systems are described in the original Millstone Unit No.
1 Station Final Safety Analysis Report Appendix B Section B.2.3 Preoperational Test Nos. 31, 32, 33, 29 and 24.
l 6.3.4.2 Reliability Tests and Inscections To assure that the ECCS will function properly if
- needed, provisions are made for testing the operability and performance i
of the various subsystems which comprise the ECCS.
This testing l
is done at a frequency that assures availability of the system.
In addition, surveillance features provide continuous monitoring of the integrity of vital portions of the ECCS.
1 Page 6.3-22 February 1987 l
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I MNPS-1 UFSAR t
Surveillance testing of the Core Spray and LPCI subsystems shall be performed as follows:
Item Frecuency i
(1)
Simulated Automatic Each Refueling i
Actuation Test Outage (2)
Pump and Valve Per Surveillance Operability Requirement 4.13 (3)
Core Spray header dp instrumentation check Once/ day calibrate Once/3 months test Once/3 months In addition to the above requirements, the LPCI pumps shall deliver at least 15,000 gpm against a system head corresponding to a reactor pressure of > 14.7 psia.
An air test shall also be performed on the drywell spray headers and no::les once every fivr years.
The FWCI system surveillance requirements are as follows:
Item Frecuency (1)
Pump and valve Per Surveillance operability Requirement 4.13 (2)
Simulated Automatic Every refueling Actuation Test-outage In addition, the quantity of water in the condensate storage tank shall be logged once a week.
The surveillance requirements for the APR system are as follows:
During each operating cycle, the following shall be performed:
(1)
A simulated automatic initiation of the system throughout its operating sequence, but excluding actual valve opening, and (2) with the reactor at low pressure, each relief valve shall be manually opened until valve operability has been verified by torus water level instrumentation, or by an audible discharge detected by an individual located outside the torus in the vicinity of each relief line.
Page 6.3-23 February 1987 P
MNPS-1 UFSAR safety / relief valve of the When it is determined that ene automatic pressure relief sucsystem is ineperable, the actuation i
lcgic of the remaining APR valves and FWCI subsystem shall ce demonstrated to be operable immediately and daily thereafter.
The surveillance requirements f0r the IC system are as fellows:
(1)
The shell side water level and temperature shall be checked daily.
(2)
Simulated autcmatic actuation and functional system testing shall be performed during each refueling cutage or wnenever ma]or repairs are completed on the system.
(3)
The system heat removal capability shall be determined once every five years.
(4)
Calibrate vent line radiation monitors quarterly.
(5)
Motor operated valves shall be tested per surveillance requirement 4.13.
The In Service Inspection and Test Program for Millstone Unit No.
1 is described in detail in the In Service Inspection and Test Ten Year Program Manual, (Reference 6.3-5, 6.3-6, 6.3-7.
and 6.3-8).
The In Service Inspec. On Program for Millstone Unit No.
I describes the items to be inspected, accessibility requirements, and the types and frequency of inspection for the ECCS systems.
6.3.5 Instrumentation Recuirements
- Details, including redundancy and logic of the emergency cere cooling system (ECCS) instrumentatien are discussed in Section 7.3.
All instrumentation required for automatic and manual initiation of the core spray system (CS), low pressure coolant injection system (LPCI),
feedwater coolant injection system (FWCI),
automatic pressure relief system (APR),
and the isolation condenser system (IC) is discussed in Section 7.3 and is designed to meet the requirements of IEEE 279 and other applicable standards.
The CS and LPCI systems are automatically initiated on low-low reactor water level or high drywell pressure.
The FWCI system
{
automatically initiates en low-low reactor water level or high drywell pressure.
The APR system is automatically initiated on high drywell pressure, reactor low-low water level after the 120 second time delay expires, and at least one core spray or low pressure coolant injection pump are running.
The IC system is automatically initiated on low-low reactor water level or reactor Page 6.3-24 December 1987
E 9
i MNPS-1 UTSAR pressure is greater than 1085 psig for 15. seconds.
The 15 secend I
+
delay prevents unnecessary system initiatien during tureine t:ne trips due to m mentary pressure sp:kes after a turbine trip.
i The FWC:, core spray, and LPCI systems autcmatically realign fr :
system flew test medes to the emergency core cooling mede cf eperation fellowing receipt of an autcmatic init:atien signal.
The core spray and LPCI systems begin inject:en into the reacter pressure vessel (RPV) when reactor vessel pressure decreases system discharge shutoff pressure.
TWCI injection begins as seen as the TWCI discharge pressure reaches a preset value (normally 1
350 psig).
i i
i i
t
?
P t
i Page 6.3-25 December 1987 i
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MNPS-1 UTSAR j
f REFERENCES i
6.3-1 Loss-of-Coolant Accident Analysts. Report for Millsecne Power Station Unit 1, NEDO-24085-1, July 1980,. w:th 1
I Addenda.
1 5.3-2
- Letter, D.C.
Switzer (NNECO),
Tc:
G.
Lear (NRC),
" Millstone Nuclear Power Station, Unit 1 Numcer 1 Loss of Coolant Accident Reanalysis Performed in Accordance with 10 CTR 50, App.
K and Revised Technical Specifications," July 9, 1975.
[
t 4
6.3-3 General Electric Standard Applicatien fer Reactor ruel, ]
l NEDE 240ll-P-A-8-US (latest approved revision).
6.3-4 Letter R.
L.
Gridley (GE) TO D.
G.
Eisenhart (NRC),
Review of Low Core flow Effects on LOCA Analysis for Cperating BWRs, September 28, 1978.
i 6.3-5 In Service Inspection Ten Year Program Manual, Oecember l
28, 1980 to December 28, 1990 Revision:
1, - dated August 31,
- 1984, Northeast Nuclear Energy Company, j
Millstone Unit No. 1.
t 6.3-6 In Service Inspection Ten Year Program Manual, October 31,
- 1984, class 2, REV 1, Northeast Nuclear Energy l
Company, Millstone Unit No. 1.
6.3-7 In Service Inspection Ten Year Program Manual, October l
31,
- 1984, Class 3, REV 1, Northeast Nuclear Energy Company, Millstone Unit No.
1.
[
I 6.3-8 In Service Test Pump and Valve Program Manual, REV 1, August 15,
- 1984, Northeast Nuclear Energy Company, Millstone Unit No. 1.
I 6.3-9 Loss-of-Coolant Accident Analynis Report for Millstene Unit 1
Nuclear Power
- Ststion, Supplement 1,
l NEDE-24085-1-P, April 1987.
-i l
i Page 6.3-26 February 1988 j
i h
1 i
i
L,--
MNPS-1 UFSAR TABLE 6.3-1
SUMMARY
DT PROVISIONS FOR EMERGENCY CORE COOLING.
1.
Loss of Normal Auxiliary Power Design Provisions
- Isolation Condenser or Feedwater' Coolant Injec-and Pressure Relief tion Subsystem and Pres-Valves sure Relief Valves 2.
Small Line Break Only (No Loss of Normal ~ Auxiliary Power)
Desion Provisions
- Any one of three condensate
- pumps, any one of three condensate booster pumps, any one of three feed pumps.
3.
Large Line Break Only (No Loss of Normal Auxiliary Power)
Desion Provisions
- Two Core Spray or One Core Spray or Two Low Pressure Subsystems (1)
Subsystem and Coolant Injection 1 Low Pressure Subsystems (3)
Coolant.Injec-tion Subsystem (2) 4.
Small Line Break Plus Loss of Normal Auxiliary Power (Standby Diesel or Gas Turbine Available)
Design Provisions
- reedwater Coolant Automatic Pressure Re-or Injection subsystem lief plus eitber Core with Gas Turbine Spray subsystem or Low Pressure Coolant Injec-tion Subsystem (with Standby Diesel) 5.
Large Line Break Plus Loss of Normal Auxiliary Power (Standby Diesel or Gas Turbine Available) fCGr.t W Mi?
.--5 e '
J, I
MNPS-1 UFSAR TABLE 6.3-1 (CONTINUED)
Desion Provisions
- T-o Core Spray or One Core Spray Subsys-Subsystems, TWCI
-tem and one Low Pressure (LPCI failure)
Coolant Injection Sub-system (diesel or- cas turbine failure) l l
e.
Post Accident Ft-covery l
l Desien Provisions
- Standby Coolant Supply System Sensit': heat is removed from the primary containment by operation of the containment cooling subsystem.
" Available alternate systenis, any one of which v'ill provide the necessary cooling f. unction.
NOTES (Item 3):
(1) Core spray subsystem consists of 1 pump plus associated equipment.
(2) LPCI subsystem consists of 2 pumps, I heat exchanger, and 2
emergency service water pumps plus associated equipment.
(3) Only 3 of 4 LPCI pumps needed.
l-Page 2 of 2 FEBRUARY 1967
MNPS-1 UTSAR TABLE 6.3-2 EMERGENCY CCRE CCOLING SYSTEM SUM.".ARY Number Cesign Effluent Required Additional of Coolant Pressure Electrical Backup runction Pumos Flow Ranoe Power Systems Core Spray 2-100%
3600 gpm 245 psi
- Normal Aux. Power 'Second C re
@ 90 psz-to or Spray Sys:er 0 psi' Emergency D:esel LPCI System Generator or Gas Turbine Generator LPCI 4-33%
7500 gpm 235 psi-Normal Aux. Pcwer Core Spray
@ 155 psi-to or System 15,000 gpm 0 psi
- Emergency Diesel 0 psia Generator or (3 pumps)
Gas Turbine Generator Normal Aux. Power Automatic TWCI:
condensate 1-100%
8000 gpm 1125 psig or Pressure Booster 1-100%
8000 gpm to Gas Turbine Relief plus reedwater 1-100%
B000 gpm 100 psig Generator Core Spray and LPCI
- Reactor vessel internal pressure to drywell pressure differential.
Automatic startup of all emergency core cooling systems is initiated by-1.
Core spray and LPCI startup on low-low water level reactor (defined as Level I, Table 7.2-1 or high pressure drywell.
l 2.
FWCI startup on low-low water level or high drywell pressure.
Page 1 of 1 December 198~
t e
MNPS-1 UFSAR TABLE 6.3-3
SUMMARY
OF CORE SPRAY SYSTEM COMPONENT l
l PERFORMANCE & DESIGN PARAMETERS
'I CORE SPPAi PL'MPS t
NUMBER 2 (ONLY l
REQUIRED TO MEET DESIGN BASIS OF NO FUEL CLAD MELT)
TYPE SINGLE-STAGE, VERTICAL, CENTRIFUGAL RATED FLOW 3600 GPM REACTOR PRESSURE AT 90 PSIG l
RATED FLCW j
SHUTOFF HEAD 255 PSIG i
i SPEED 3600 RPM l
SEALS MECHANICAI, DRIVE ELECTRICAL MOTOR POWER SOURCE NORMAL AUXILIARY OR STANDY DIESEL CR
,q GAS TURBINM PUMP CASING CAST STEEL IMPELLER BRONZE i
SHAFT STAINLESS STEEL l
1 POWER @ MAX. OPERATING 800 HP CONDITICNS NPSH (AVAILABLE)
DESIGN PRESSURE 350 PSIG DESIGN TEMPERATURE 3500 t
F CODE ASME SECTION III, CLASS "C"
TRIP DELAY (SIGNAL TO 49 SECONDS WITH DIESEL POWER, 55 RATED FLOW)
SECONDS WITH GAS TURBINE j
t PAGE 1 OF 2 FEBRUARY 1987 s
P
l l
MNPS-1 UFSAR TABLE'6.3-3 (CONTINUED)
I 1
_.3.._
._ _.... a.
l CODE USAS B31.1 DES:GN P? ESSURE / TEMP 0 F SUCT:CN PIP!NG 150 PSIG/200 t
0 DISCHARGE PIPING 350 PSIG/350 F
1 i
SPRAY-HEADERS 1
NUMBER 2
t NUMSER OF FLOW RIFES 57 PER HEADER i ALTERNATING PATTERN
[
NUMBER OF NOZZLES 57 PER HEADER i ALTERNATING PATTERN 4
TYPE CF NOZZLES 3/4 FULLJET - STAINLESS STEEL (304) i i
i SPARGER MATERIAL 304 S.S.
1 DESIG:. CODE ASME, SECTION.III, CLASS "B"
i
.I i
C i
I i
i r
r t
8 i
PAGE 2 OF 2 FEBRUARY 1987 l
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MNPS-1 UFSAR l
i TABLE 6.3-4 I
f
SUMMARY
OF LOW PRESSURE COOLANT INJECTION j
CCMPONENT PERFORMANCE & DESIGN PARAMETERS l
L?C: PUMPS NUMBER 4 ( 3 REQUIRED TO MEET DES:GN BAS!S )
TYPE SINGLE STAGE I
SEALS MECHANICAL I
DRIVE ELECTRIC MOTOR, 4160V, 30 500 BHP l
POWER SOURCE NORMAL AUXIL!ARY OR STANDBY DIESF' OR GAS TURBINE GENERATCR SPEED 3600 RPM i
i SHUTOFF HEAD 625 FT.
i PUMP CASING CAST STEEL IMPELLER BRONZE SHAFT STAINLESF STEEL l
CODE ASME SECTION III, CLASS "C"
i 0
DESIGN PRESS / TEMP 400 PSIG/350 F
.i PERFORMANCE CHARACTERISTICS-3 PUMPS RUNNING l
AT 9 PSI REACTOR PRESSURE FLOW 5000 GPM EACH, 15,000 GPM TOTAL HEAD 240 FEET POWER 527 HP EACH, 1581 HP TOTAL l
NPSH (AVAILABLE) 33 FEET PAGE 1 OF 3 I
FEBRUARY 1937
e
=
(
MNPS-1 UFSAR
[
TABLE 6.3-4 (CCNTINUED)
{
i AT 165 PSI REACTOR PRESSURE FLOW 2500 GPM EACH, 7530 GPM TOTAL HEAD 475 FEET POWER 440 HP EACH, 1320 HP TOTAL NPSH (AVAILABLE) 42 FEET i
TRIP DELAY SIGNAL TO RATEIP 40 SECONDS 4:TM DIESEL POWER, 75 SECCNDS WITH GAS TURBINE PUMPS. EMEPGENCY SERVICE '4ATER NUMBER 4
(2 REQUIRED TO PROVIDE COOLING CAPACITY)
TYPE VERTICAL, CENTRIFUGAL DRIVE ELECTRIC MOTOR, 4160V, 30, 400 BHP POWER SOURCE
. AUXILIARY TRANSFORMER OR STANDBY DIESEL OR GAS TURBINE i
SPEED 1770 RPM RATED CAPACITY 2500 GPM EACH, 5000 GPM TOTAL RATED HEAD (APPROXIMATELY) 425 FEET i
- SHUTOFF HEAO 700 FEET l
LPCI HEAT EXCHANGERS NUMBER 2 (1 REQUIRED)
SHELL SIDE TUBE SIDE F
l DESIGN TEMPERATURE 2050 F
2050 DESIGN PRESSURE 300 PSIG 300 PSIG FLUID CIRCULATED DEMIN. WATER EMERG. SERVICE WATER FLOW RATE 5000 GPM 5000 GPM i
INLET TEMPERATURE 165 F
750 F
OUTLET TEMPERATURE 1490 F
91 F
]
FOULING FACTOR
.0005
.0005 HEAT EXCHANGEF DUTY 40,000,000 Btu /Hr each PAGE 2 OF 3 FEBRUARY 1997
a
~ ~ -.
.. ~.
.~.
I i
MNPS-1 UFSAR TABLE 6.3-4
[
(CCNT!NUED)
!.i CONTAINVENT SPRAY SPARGERS t
ORYWELL I
FLCW PER HEADER 6175 GPM NOZZLES PER HEADER 104 NOZZLE TYPE FOGJET NOZZLES PIPE SIZE 8" XX SUPPRESSICN CHAMBER NO CF HEADERS 1
FLCW PER HEADER 325 GPM NOZZLES PER HEADER 16 i
NOZZLE TYPE FOGJET NOZZLES PIPE SIZE 4" SCH. 80 2i l
i
'i i
3
~!
~
l d
I l
2 PAGE 3 OF 3 FEBRUARY 1987 l
l
. = -...... -
4 e
i a
i MNPS-1 UFSAR f
TABLE 6.3-5 1
SUMMARY
CF FEEDWATER COOLANT INJECTICN
)
SYSTEM COMPONENT PERFORMANCE i DESIGN PARAMETERS l
CONOENSATE FUuPS I
NUMBER 3
TYPE VERTICAL RATED FLOW B000 GPM RATED HEAD 375 FT.
i SPEED 1200 RPM j
DRIVE ELECTRICAL MOTOR 900 BHP TRIP DELAY SEE REACTOR FEED CONDENSATE BOOSTER PUMPS-i NUMBER 3
TYPE HORIZONTAL, CENTRIFUGAL j
i RATED FLOW 8000 GPM RATE HEAD B30 FEET SPEED 1800 RPM DRIVE ELECTRICAL MOTOR, 2000 BHP REACTOR FEED PUMP l
NUMBER 3
l TYPE HORIZONTAL, CENTRIFUGAL RATED FLOW 8000 GPM l
PAGE 1 OF 2 FEBRUARY 1987
~.
, e l
MNPS-1 UFSAR i
TABLE 6.3-5 (CONTINUED) 1 RATED HEAD 2900 FEET SPEED 3600 RPM l
i DRIVE ELECTRICAL MOTOR, 7000 BHP TRIP DELAY l
(SIGNAL TO RATED FLOW) 90 SECONDS EMERGENCY CCNDENSATE TRANSFER PUMP l
l NUMBER 1
TYPE CENTRIFUGAL SINGLE STAGE l
RATED FLOW 3600 GPM RATED HEAD 100 PSIG SPEED DRIVE ELECTRICAL MOTOR
{
MAXIMUM DELAY TIME-TRANSFER PUMP START 60 SECONDS CONDENSATE HOT WELL MINIMUM VOLUME 75000 GALLONS l
CONDENSATE STORAGE TANK MINIMUM VOLUME 225,000 GALLONS'
'I
?-i i
i l
PAGE 2 OF 2 j
FEBRUARY 1997 4
I i
r
MNPS-1 UFSAR i
TABLE 6.3-6 l
SUMP.ARY OF AUTCMATIC PRESSURE RELIEF (APR) SYSTEM l
PERFCRMANCE & DESIGN PARAMETERS NUMBER OF APR DESIGNATED 4
SAFETY RELIEF VALVES VALVE TYPE PILOT OPERATED SAFETY / RELIEF VALVE SET PRESSURE 1095 To 1125 PSIG CAPACITY 800,000 LBM/HR/ VALVE-SATURATED STEAM OPERATOR ACTUATICN PNEUMATIC (NITROGEN GAS) SUPPLIED BY DRYWELL ATMOSPHERIC CCMPRESSOR (100 PSIG)
PNEUMATIC SOLENOID POWER SUPPLY -
125 VDC.
'SEE P':EUMATIC -
ACCUMULATORS TRIP DELAY 120 SECONDS DELAY FROM SIGNAL TO ACTUATION TO PROVIDE TIME. FOR OPERATOR INTERVENTION 4
['NEUMATIC ACCUMULATORS f
CAPACITY SUFFICIENT TO PROVIDE FOR FIVE RELIEF VALVE ACTUATIONS IN THE EVENT i
OF LOSS OF PNEUMATIC SUPPLY FROM DRYWELL ATMOSPHERIC COMPRESSOR l
DISCHARGE VACUUM BREAKERS NUMBERS OF VACUUM BREAKERS j
PER-RELIEF VALVE DISCHARGE j
LINE 3
~!:!
)
PAGE 1 OF 1 FEBRUARY 1987 j
i
_m i
s'
',4 9
l MNPS-I UFSAR i
i TABLE 6.3-7 j
l
SUMMARY
OF ISOLATION CONDENSER SYSTEM PERFORMANCE & DESIGN PARAMETERS t
ISOLAT:CN CONDENSER RATFD CAPACITY 206 X 106 BTU /HR (3% REACTOR POWERI' I
DURATION OF OPERATION WITHOUT SHELL SIDE MAKEUP l
WATER APPROX. 30 MIN.
i ELEVATICN OF TUBE BUNDLES ABOVE NCRMAL REACTOR WATER e'
LEVEL 25 FT NCMINA' NORMAL SHELL MAKE UP SOURCE FIRE MAIN
-l BACKUP SHELL MAKE UP SOURCE CONDENSATE TRANSFER SYSTEM INITIAL SHELL WATER CLEAN DEMINERALIZED WATER VIA
-f CONDENSATE TRANSFER l
POWER SOURCE FOR CONDENSER l
RETURN VALVE 125 VOLT DC j
SHELL SIDE j
MATERIAL CARBON STEEL l
DESIGN PRESSURE 5 PSIG MAX.
d CAPACITY 15,500 GALLONS, NORMAL 4
TUBE SIDE J
MATERIAL INCONEL 600 DESIGN PRESSURE 1250 PSIG I
1 PAGE 1 OF 1 FEBRUARY 1987 1
. ~
i e,*
f i
}
MNPS-1 UFSAR f
TABLE 6.3-3
SUMMARY
OF RESULTS OF LOCA ANALYSIS-o BREAK SIZE CORE-WIDE j
o LOCATION PEAK LOCAL METAL-WATER
-l o SINGLE FAILURE PCT ( OH OXIDATION ( *. )
REACTION fi) o 4.3 FT (DBA) 2200 (1) 8.2 c.19-2 o RECIRC SUCTION o
LPCI INJECTION VALVE l
2 o
0.10 FT 2155 (1)
(2)
(2) o RECIRC DISCHARGE o GAS TURBINE 1
(2)
PCT FROM CHASTE
( )
LESS THAN DBA VALUE l
3 i
t
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t ftervw v i 4Y7 i
pw, 3 n, i
a
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.,, 4 l
1 MNPS-1 UTSAR i
l TABLE 6.3-9 MAPLHGR.VERSUS AVERAGE PLANAR EXPOSURE TUEL TYPE:
?9DRB282
-1 l
AVERAGE PLANAR EXPOSURE MAPLHGR*
PCT OXICATICN (mwd /st)
(KW/ft)
('T)
TRACT:CN i'
200 10.7 2198 0.036 c
1000 10.7 2197 0.035 5000 11.1 2198 0.034 10000 11.3 2197 0.033 15000 11.3 2199 0.033 20000 11.2 2196 0.033 25000 11.2 2198 0.033 30000 11.0 2198 0.083 35000 10.4 2084 0.057 40000 9.8 1976 0.038 i
l TUEL TYPE:
PSDRB283H
?
AVERAGE PLANAR l
EXPOSURE MAPLHGR*
PCT OXIDATION (mwd /st)
(KW/ft)
(*T)
TRACTION 200 10.7 2198 0.036 1000 10.8 2198 0.035 5000 11.1 2200 0.034 10000 11.3 2200 0.333 15000 11.3-2199 0.033 20000 11.2 2199 0.033 i
25000 11.0 2200 0.077 i
30000 11.0 2199 0.085 35000 10.9 2185 0.079 40000 10.4 2096 0.060-45000 10.0 2015 0.046 1
.i Page 1 of 2 February 1987 i
i
.l
o....
MNPS-1 UTSAR TABLE 6.3-9 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE (CONTINUED)
^
TUEL TYPE:
SPBDRB300 AVERAGE PLANAR EXPOSURE MAPLHGR*
PCT OXIDATION (mwd /st)
(KW/ft)
(*T)
TRACTION 200 10.8
.2200 0.036 1000 10.8 2195 0.035 5000 11.2 2199 0.034 10000 11.3 2198 0.033 15000 11.3 2200 0.033 20000 11.2 2197 0.033-25000 11.1 2194 0.077 35000 10.3 2074 0.056 45000 9.1 1867 0.023
- A MULTIPLIER OT 0.95 IS TO BE APPLIED TO THE VALUES IN THE ABOVE TABLES WHEN OPERATING AT LESS THAN 90% OT RATED CORE y
TLOW.
?
TUEL TYPE:
BD338A (GE-8B)
AVERAGE PLANAR MAPLHGR (kW/Tt) l EXPOSURE MOST LEAST OXI?ATION (GWD/ST)
LIMITING
- LIMITING
(*T)**
TRACTION **
l 0.2 10.38 10.70 2076 0.065 l
1.0 10.55 10.80 2107 0.070 l
5.0 11.40 11.40 2195 0.085 1
10.0 11.50 11.50 2199 0.083 l
15.0 11.50 11.50 2198 0.083 l
20.0 11.40 11.40 2194 0.081 4
25.0 11.04 11.06 2197 0.082 i
35.0 9.24 9.27 2058 0.053 45.0 7.44 7.47 1822 0.020 50.0 6.54 6.58 1672 0.005
- Enriched lattices only.
Natural lattices are non-limiting for LOCA events.
I
- Maximum lattice values.
Values for all lattices contained in Reference 6.3-9.
Page 2 of 2 rettuary 1988 t
o a>
MNPS-1 UPSAR
\\
TABLE 6.3-10 I
ECCS LOAD TIME SEQUENCE, GAS TURBINE (PWCI AND HALP i
OF ECCS LCADS) DIESEL GENERATOR (NORMAL ECOS) LCADING SEOUENCE j
CAS DIESEL.
TUREINE (SECON05)
ISECCNOS?
ACCIOENT.
0 0
l SIGNAL TO START DIESEL AND GAS TURBINE.
3 3
DIESEL GENERATOR LCADING STARTS.
START PIRST 13 LPCI PUMP.
START RECIRCULATION VALVE CLOSURE.
FIRST LPCI PUMP AT RATED SPEED. START SECOND 18 LPCI PUMP.
LPCI INJECTION VALVES START OPENING (IP LOW 22 PRESSURE PERMISSIVE).
SECOND LPCI PUMP AT RATED SPEED.
23 PIRST CORE SPRAY PUMP STARTS.
23 PIRST CORE SPRAY PUMP AT RATED SPEED. CORE SPRAY 23 INJECTION VALVE STARTS TO OPEN (IP LOW PRESSURE PERMISSIVE.
CORE SPRAY INJECTION VALVE OPEN.
38 LPCI INJECTION VALVE OPEN.
43 RECIRCULATION LINE VALVES ARE CLOSED.
43 GAS TURBINE READY TO LOAD.
CONDENSATE PUMP, 48 2ND CORE SPRAY PUMP, 3RD AND 4TH LPCI PUMPS ALL START.
MOTOR-OPERATED BLOCK VALVE STARTS TO OPEN.
CONDENSATE PUMP AT RATED SPEED.
START CONDENSATE 51 BOOSTER PUMP.
CONDENSATE BOOSTER PUMP AT RATED SPEED.
START 54 PEEDWATER PUMP
)
Page 1 of 2 March 1989 i
l
MNPS-1 UTSAR TABLE 6.3-10 ECCS LCAD TIME SEQUENCE, GAS TURBINE (FWCI AND HALF CT ECOS LCADS) DIESEL GENERATCR (NORMAL ECCS) LCADING SEOUENCE (CONTINUED)
GAS DIESEL TURBINE (SECONDS)
(SECCNDS)
CORE SPRAY INJECTION VALVE FULLY CPEN.
58 FEEDWATER PUMP AT RATED HEAD.
START TO OPEN 64 FEEDWATER REG VALVE.
LPCI INJECTION VALVE TULLY CPEN.
66 FWCI AT RATED FLCW.
90
[
?
v i
Page 2 of 2 March 1989
- e MNPS-1 UTSAR TABLE 6.3-11 SIGNITICANT INPUT VARIABLES USED IN LOSS-0T-COOLANT ACCIDENT ANALYSIS variable value A.
PLANT PARAMETERS CORE THERMAL POWER 2051 MWt (102% of rated) 6 VESSEL STEAM OUTPUT 8.15 x 10 lbm/hr VESSEL STEAM DOME PRESSURE 1035 psig RECIRCULATION LINE BREAK AREA 4.3 ft (DBA) l FOR LARGE BREAKS B.
EMERGENCY CORE COOLING SYSTEM PARAMETERS 1
LOW PRESSURE COOLANT INJECTION SYSTEM VESSEL PRESSURE AT WHICH FLOW MAY COMMENCE 235 PSIG TOTAL MINIMUM RATED FLOW AT 0 PSIG VESSEL PRESSURE 3 PUMPS 15000 GPM INITIATING SIGNALS (1)
LOW-LOW WATER LEVEL, OR 78-1/2" ABOVE TOP OF ACTIVE FUEL HIGH DRYWELL PRESSURE 2 PSIG MAXIMUM ALLOWABLE TIME DELAY 23 SECONDS - DIESEL POWER TROM INITIATING SIGNAL TO 53 SECONDS - GAS TURBINE POWER l
PUMPS AT RATED SPEED INJECTION VALVE FULLY OPEN 40 SECONDS-DIESEL POWER 66 SECONDS-GAS TURBINE POWER Page 1 of 3 February 1988 l
MNPS-1 UTSAR TABLE 6.3-11 (CONTINUED) i Variable Value CORE SPRAY SYSTEM VESSEL PRESSURE AT WHICH TLOW MAY COMMENCE 255 PSID (VESSEL TO DRYWEL* )
MINIMUM RATED TLCW, AT VESSEL 3600 GPM AT 90 PSID (VESSEL TO DRYWELL)
PRESSURE INITIATING SIGNALS (1)
LOW-LOW WATER LEVEL, OR 78-1/2" ABOVE TOP OF ACTIVE l
TUEL HIGH DRYWELL PRESSURE 2.0 PSIG MAXIMUM ALLOWED (RUNOUT) TLOW 5300 GPM MAXIMUM ALLOWED DELAY TIME TROM 28 SECONDS - DIESEL POWER FROM INITIATING SIGNAL TO PUMP 53 SECONDS - GAS TURBINE POWER AT RATED SPEED DIESEL POWER INJECTION VALVE FULLY OPEN 49 SECONDS 58 SECONDS - GAS TURBINE POWER NUMBER Or APR VALVES 4
TOTAL MINIMUM TLOW CAPACITY, AT 3,200,000 lbm/hr A VESSEL DOME PRESSURE OF 1035 PSIG INITIATING SIGNALS (1)
LOW-LOW WATER LEVEL, AND 78.5" ABOVE TOP OF ACTIVE TUEL HIGH DRYWELL PRESSURE. AND
> 2 PSIG i
(1)
THIS ANALYSIS IS BOUNDING FOR INITIATING SIGNALS WITHIN THE l
INDICATED RANGE.
Page 2 of 3 February 1988 l
w--
~
e..4 i
l MNPS-1 UFSAR TABLE 6.3-11 (CCNTINUED) value Variable I
SIGNAL THAT AT LEAST CNE 100 PSIG LOW PRESSURE ECCS PUMP IS l
RUNNING (PUMP DISCHARGE PRESSURE); AND DELAY TIME FROM ALL INITIATING SIGNALS 120 SECONDS COMPLETED TO THE TIME-VALVES ARE OPEN C.
FUEL PARAMETERS (REF. 6.3-1)
RELOAD FUEL TYPE FUEL BUNDLE GEOMETRY 8X 8 D
LATTICE NUMBER OF FUELED RODS PER 62 BUNDLE PEAK TECHNICAL SPECIFICATION 13.4 KW/FT. (GE-78)
LINEAR HEAT GENERATION RATE 14.4 KW/FT. (GE-8B)
INITIAL MINIMUM CRITICAL POWER 1.24 RATIO
- DESIGN' AXIAL PEAKING FACTOR 1.57 i
^
TO ACCOUNT FOR THE 2's UNCERTAINTY IN BUNDLE POWER REQUIRED BY a
APPENDIX K.
THE SCAT CALCULATION IS PERFORMED WITH AN MCPR OF 1.22 (I.E., 1.24 DIVIDED BY 1.02) FOR A BUNDLE WITH AN INITIAL e
MCPR OF 1.24.
Page 3 of 3 February 1988 i
i
l i
Docket No. 50-245 B14692 i
t
-) - Exhibit 9 Millstone Nuclear Power Station, Unit No. ]
s i
t I
i 1.!
1 e
l i
l December 1993
- e t
53rI INSPECTION CBSERVATION NUMBER 1-16-88 CATEGORY B T
INSPECTOR:
Sterner 0 ATE /T ME:
10/17/88 1100 WORK PLAN SECTION:
Engineering / Design i
Issue:
Lacx o:
following:
Information suppor:Ang tne cesign oasts of tne
{
1)
FWCI flow rate (8000 gpa) 2)
Time to rated rWCI flow after LNP (90 sec) 3)
ECT pump flow rate (3600 gps) 4)
Tl:e to ECT pump start (60 see) 5)
Min CST volume for rWCI operation (225,000 gal) (or 250,000 gallen T.S. Amendment 18) i DISCUSSION:
relate to post-LOCA TWCI operation) Calculations supporting Onese paramet twnten are not available.
Therefore, identified. the basis of these design values
- can not be
- (Reference UFSAR Table 6.3-5, Tech.
i spec. 3.5.C.2)
RESOLUTION:
The Observation 1-16-66 information supporting the design basis of various rWCI related pointed out tne lacx of-parameters.
and your consideration.The following comments are offered for perspective i
1.
rWCI riowrate of 8000 gpa - The TWCI 1
based on the original design flow of a single condensate /flowrate of 8000i b
i
?
F 5
s 3
e
feedwater train.
The 8000 gpa is not based on any safety system requirement and no calculations extst Vith respect to plant performance with the use of rWCI versus LPCI/CS.
The Millstone Unit one safety analyses have always snown that the more limiting cases have been those where the TWCI is either not available or not utill:ed.
This has resulted in a requirement that the system be available but no explicia minimum flow requirement has ever been identified.
r 2.
The time to rated TWCI flow after an LNP (90 see) - This I
nuater appears to have been chosen to assure that adequate l
time exists for gas turbine start and rWCI string startup.
i No explicit requirement for the 90 seconds has been ident:fied, and the value appears to be arbitrary.
3.
Emergency condensate Transfer pump flowrate (3600 GPM) - The
)
value appears to have no basis.
I The apparent lack of or inconsistent design bases for the
]
Millstone Unit 1 feedwater/ condensate system has caused a series of questions to be raised concerning safety, the need to reconstruct the design bases and how the design process is controlled to assure that the bases are understood when changes
-l are made.
The safety concern has been answered, which leaves the second two concerns.
The present condensate /feedwater system although not well documented, performs it primary functions as witnessed by the years of successful operation.
At this time, it is not NU's intent to reconstruct all design bases for the unit.
There is a l
significant project underway for all units (Design Basis Reconstruction) to identify and document the design bases for most of the major systems and components.
This project will document the design bases when available but will not reconstruct the bases.
The systems and components that lack design bases j
will be identified and if significant, a request will be initiated to reconstruct the bases or resolve the issue that was presented.
This approach to reconstruction is prudent and has j
been acknowledged by both NU and the regulatory agencies.
l The design change process at NU assures appropriate reconstruction of design bases on an as needed basis.
This is assured in both NEO 3.03, " Preparation, Review, and Disposition
)
of Plant Design Change Records" (Step 6.2.3.1) and in NEO 3.12, I
" Safety Evaluations" (Step 6.3.4).
Both procedures require that the design bases be understood prior to making the change and i
that the differences in the change be documented.
NEO 3.03 also requires updates of appropriate plant design documents to reflect the changes.
These "in place" procedures should assure that the bases for any system or component are understood and documented prior to a change being made.
I i
~
a i
i The condensate /feedwater system at Millstone Unit I was originally envisioned as a 2/2/2 pump conYiguration but has been run as a 3/3/2 pump configuration.
As the change to the 3/3/2 configuration was done and accepted early in plant life, this constitutes the design bases and the plant documentation should have been changed to reflect the change at the time.
Future-
~
as they occur, should be performed consistent with the
- nanges, NEO procedures thus too changes to the bases and documentation of 2.
j; these changes should occur.
(-
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it(.
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2 hru 7d
.es
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- 9M.
- sW i
.:st
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- adC upt1
- k
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114 k
i v.
1
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Docket No. 50-245 B14692 J
t t
-I
?
f t
t t
l
., - Exhibit 10 Millstone Nuclear Power Station, Unit No.1 i
i l
1
)
1 December 1993 i
'