ML17059B096
| ML17059B096 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 04/02/1996 |
| From: | Hood D NRC (Affiliation Not Assigned) |
| To: | Sylvia B NIAGARA MOHAWK POWER CORP. |
| Shared Package | |
| ML17059B097 | List: |
| References | |
| GL-88-10, TAC-M74436, NUDOCS 9604080018 | |
| Download: ML17059B096 (22) | |
Text
gPR "50Ij~
~o Cy OO 0
Yg 0
~O
++*++
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&0001 April 2, 1996 Qo Zz0 Hr. B. Ralph Sylvia Executive Vice President, Nuclear Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station P. 0.
Box 63
- Lycoming, NY 13093
SUBJECT:
NRC STAFF'S EVALUATION OF THE NINE MILE POINT NUCLEAR STATION UNIT NO.
1 INDIVIDUALPLANT EXAMINATION (IPE)
SUBMITTAL (TAC NO. M74436)
Dear Hr. Sylvia:
By letter dated July 27,
- 1993, as supplemented June 26, 1995, you responded to Generic Letter (GL) 88-20, "Individual Plant Examinations for Severe Accident Vulnerabilities," and Supplements 1, 2, and 3, thereto.
Mith the assistance of our contractors, we have completed our review of the IPE submittal for internal events.
The evaluation package consists of:
~
The Staff Evaluation Report (SER)
(Enclosure 1)
The contractor's Technical Evaluation Reports (TERs) for the front-end, back-end, and human reliability analysis reviews (Enclosures 2, 3, and 4)
The Nine Mile Point Unit 1 submittal did not identify any severe accident vulnerabilities associated with either core damage or poor containment performance.
Me noted that as a result of the IPE or other industry initiatives, you implemented several procedural enhancements and hardware modifications which were reflected in your core damage frequency (CDF) estimate of 5.5E-6 per reactor year from internally initiated events, excluding internal flooding.
Based on our review of the Nine Mile Point Nuclear Station Unit No.
1 IPE submittal and associated documentation, we conclude that you have fully met the intent of GL 88-20.
GL 88-20 suggested that licensees could use their IPE submittals to address, among other safety issues, Unresolved Safety Issue (USI) A-45, "Shutdown Decay Heat Removal Requirements."
In your response to GL 91-06, you had proposed to use the IPE submittal to i espond to Generic Issue A-30, "Adequacy of Safety-ha ~
9604080018';9'60402 PDR ADQCK'5000220 P
'>PDR)
aQ gs
B. Sylvia r
t Related DC Power Supplies."
These two generic issues are adequately resolyed,"
for Nine Mile Point Nuclear Station Unit No.
1 by your 'IPE,s'ubm'ittal.
If you have any comments or questions, please contact me at (301) 415-3049.
\\
Sincerely, i
Original signed by:
11 1
Darl S.
Hood, Senior'roject 'Manag'er Project Directorate I-l Division of Reactor Projects -/II-
'ffice of Nuclear Reactor Regulation Docket No. 50-220
Enclosures:
1.
Staff Evaluation 2.
TER (Front-End) 3.
TER (Back-End) 4.
TER (Human Reliability Analysis) cc w/encl 1:
See next page i1ti1 w Enclosures 1-4 Docket File PUBLIC DHood
- BNorris, SRI ACRS w Enclosure 1 onl PDI-1 Reading SVarga JZwolinski SShankman SLittle DClark OGC CCowgill, RGN-I RHernan EButcher
- ELois, RES
- MDrovin, RES EKelly, RI DOCUMENT NAME:
G: iNMPIiNM174436. I PE To receive e copy of this document, indicete in the box: "C" ~ Copy without enciosures "E" ~ Copy with en osures "N" ~ No copy OFFICE LA:PDI-PM: DRPE w
PM: PD I-1 D:PD NAME DATE SL'ttl N/ t /96 RClark:sm 03
/96 DHood
/96 OFFICIAL RECORD COPY SSha
/96 03/
/96
I I I
. ~
k I
I.
I tt I
ter l
I
'I i(
kt I
I 11 I
F, p J1 r II I
J
- C I
l
'I I
J I
,r,k lt
~
j I
11
'k
,I
'Itk lk i
I' J
I I,
gV'J i
p-"
I t,'
I II tt r.
i IIJ
B. Sylvia Related DC Power Supplies."
These two generic issues are adequately resolved for Nine Mile.Point Nuclear Station Unit No.
1 by your IPE submittal.
If you have any comments or questions, please contact me at (301) 415-3049.
Sincerely, Original signed by:
Docket No. 50-220 Darl S.
Hood, Senior Project Manager Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation I
Enclosures:
1.
Staff Evaluation 2.
TER (Front-End) 3.
TER (Back-End) 4.
TER (Human Reliability Analysis) cc w/encl 1:
See next page Distribution:
w Enclosures 1-4 Docket File PUBLIC DHood
- BNorris, SRI ACRS Enclos re on PDI-1 Reading SVarga JZwolinski SShankman SLittle DClark OGC CCowgill, RGN-I RHernan EButcher
- ELois, RES
- MDrovin, RES EKelly, RI DOCUMENT NAME:
G:iNMPliNM174436. IPE To receive e copy of this document. indicate In the box:
C" ~ Copy without enciosures "E
~ Copy with en osures "N
~ No copy OFFICE LA:PDI-PH: DRPE w
PM: PDI-1 D:PD NAME SL'ttl
=
RClark:s DHood SSha DATE (8/ I /96 03
/96
/96 OFFIC AL RECORD COPY
/96 03/
/96
B. Sylvia Related DC Power Supplies."
These two generic issues are adequately resolved for Nine Mile Point Nuclear Station Unit No.
1 by your IPE submittal.
If you have any comments or questions, please contact me at (301) 415-3049.
Sincerely, Docket No. 50-220 OP /
Darl S.
Hoo
, Senior Project Hanager Project Directorate I-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosures:
1.
Staff Evaluation 2.
TER (Front-End) 3.
TER (Back-End) 4.
TER (Human Reliability Analysis) cc w/encl 1:
See next page
V 1
'J sp 4
B. Ralph Sylvia Niagara Mohawk Power Corporation Nine Hile Point Nuclear Station Unit No.
1 CC:
Hark J. Wetterhahn,. Esquire Winston 5 Strawn 1400 L Street, NW Washington, DC 20005-3502 Supervisor Town of Scriba Route 8, Box 382
- Oswego, NY 13126 Hr. Richard B. Abbott Vice President Nuclear Generation Niagara Mohawk Power Corporation Nine Nile Point Nuclear Station P.O.
Box 63
- Lycoming, NY 13093 Resident Inspector U.S. Nuclear Regulatory Commission P.O.
Box 126
- Lycoming, NY 13093 Gary D. Wilson, Esquire Niagara Mohawk Power Corporation 300 Erie Boulevard West
- Syracuse, NY 13202 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. F. William Valentino, President New York State
- Energy, Research, and Development Authority 2 Rockefeller Plaza
- Albany, NY 12223-1253 Mr. Norman L. Rademacher Unit 1 Plant Manager Nine Mile Point Nuclear Station P.O.
Box 63
- Lycoming, NY 13093 Hs. Denise J. Wolniak Manager Licensing Niagara Mohawk Power Corporation Nine Hile Point Nuclear Station P.O.
Box 63
- Lycoming, NY 13093 Charles Donaldson, Esquire Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D.
Eddy State of New York Department of Public Service
-~
Power Division, System Operations 3 Empire State Plaza
- Albany, NY 12223 Hr. Hartin J.
HcCormick, Jr.
Vice President Nuclear Safety Assessment and Support Niagara Mohawk Power Corporation Nine Mile Point Nuclear Station P.O.
Box 63
- Lycoming, NY 13093
NINE MILE POINT NUCLEAR STATION UNIT NO. I INDIVIDUALPLANT EXAMINATION STAFF EVALUATION REPORT Enclosure I
'I I C}
r,
>C pk
1.0 On July 27, 1993, Niagara Hohawk Power Corporation (NHPC or the licensee) submitted the results of the individual plant examination (IPE) for Nine Hile Point Nuclear Station Unit No.
1 (NHP1) in response to Generic Letter (GL) 88-20 and associated supplements.
On April 21, 1995, the NRC staff sent questions to the licensee requesting additional information.
The licensee responded in a letter dated June 26, 1995.
A "Step 1" review of the NHP1 IPE submittal was performed and involved the efforts of Science 8 Engineering Associates, Inc.
(SEA), Scientech, Inc.,
and Concord Associates in the front-end, back-end, and human reliability analyses (HRA), respectively.
The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities.
Therefore, the review considered (1) the completeness of the information and (2) the reasonableness of the results given the design, operation, and history of NHPl.
A more detailed review, a "Step 2" review, was not performed.
A summary of contractors'indings and the staff's evaluation is provided below.
Details of the contractors'indings are in the technical evaluation reports (Enclosures 2, 3, and 4) attached to this staff evaluation report (SER).
In accordance with GL 88-20, the licensee proposed to resolve Unresolved Safety Issue (USI) A-45, "Shutdown Decay Heat Removal
[DHR] Requirements."
The licensee also proposed to resolve GI A-30, "Adequacy of Safety Related DC'ower Supplies,"
as part of the NHPl IPE.
No other specific USIs or GIs were proposed for resolution.
2.0 A
ON The NHP1 plant is a boiling-water reactor (BWR) 2 with a Hark I containment.
In the IPE, the licensee has estimated a total core damage frequency (CDF) of 5.5E-6/reactor year excluding flooding.
This CDF is lower than the CDFs for most BWRs.
It appears that this low CDF estimate is driven by a low relative CDF contribution of transient events (8E-7/reactor year); the average transient relative CDF for BWRs is 8E-6.
Station blackout contributes 64X transients 14X, loss-of-coolant accidents (LOCAs) 13X, and anticipated transients without scram (ATWS) 10X.
Internal flooding was screened from the analysis on the basis of semi-quantitative flood scenario evaluations.
The important system/equipment contributors to the estimated CDF that appear in the top sequences are:
failure to recover ac power, loss of ac and dc power, failure of the electromatic relief valves to reclose, and failure of the diesel fire pump to supply the reactor pressure vessel.
It appears that the significant initiating events and dominant accident sequences were examined in the NHPl Level 1 analysis.
Based on the licensee's IPE process used to search for DHR vulner abilities and a review of NHP1 plant-specific features, the staff finds the licensee's DHR evaluation consistent with the intent of the resolution of USI A-45.
The licensee performed an HRA to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure events.
The licensee identified the following operator actions as important in the estimate of the CDF:
ac power recovery, emergency diesel load shedding
fjy k) 1 t
C\\
1 I
under LOCA conditions, reactor pressure vessel depressurization, prevention of the emergency
(.isolation) condenser (EC) isolation and EC recovery after isolation, core spray injection permissive calibration, feedwater control given loss of instrument air, dc load shedding given station blackout, and containment spray alignment for torus cooling mode.
The licensee evaluated and quantified the results of the severe accident progression thr ough the use of a containment event tree and considered uncertainties in containment response through the use of sensitivity analyses.
Early releases (in less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from accident initiation) occur 26X of the time, intermediate (between 6 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 48X of the time, and late (after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 14X of the time; the containment remains intact 13X of the time.
The staff noted that the licensee's definition of early containment failure (in less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from accident initiation) was different from the typical definition of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from accident initiation.
The licensee considered large releases as an indicator of containment performance.
Large releases are defined in terms of accident sequences that will result in high releases (greater than 10X in CsI fission products) and early releases.
Large release represents 13X of the total release frequency and is dominated by wetwell overpressure mainly occurring in ATWS sequences.
The staff noted that although ATWS is not an important contributor to the overall CDF (10X), it is an important contributor to large releases.
It appears that the important severe accident phenomena were considered in the NHP1 Level 2 analysis.
The licensee's response to containment performance improvement (CPI) program recommendations is consistent with the intent of GL 88-20 and associated Supplement 3.
Some insights and unique plant safety features identified at NHPI are:
l.
The ECs do not initially require electrical power to provide core cooling thus extending the time for AC recovery during station blackout.
2.
A hardened containment vent provides a backup to loss of containment cooling.
3.
Eight-hour battery lifetime is relatively long compared to battery lifetimes at other BWRs and increases the likelihood of recovering offsite
.power.
4.
The diesel-driven firewater pump provides makeup to the,ECs; and 5.
The capability to power the control rod drive pumps with the diesel generators ensures an additional source of makeup to the vessel even if offsite power is lost.
The licensee did not define what constitutes a plant vulnerability to severe accidents.
It is stated in its submittal that no unusual or unique contributors to core damage nor unusually poor containment performance have been identified.
However, the licensee implemented the flowing improvements:
I
d' 1.
2.
Revision 4 of the BWR Owners Group emergency procedure guidelines; and 3.
Initiation of the use of cross-tie containment spray raw water to core spray as an alternative source of injection to the reactor pressure vessel; initiation of the option align the containment spray raw water to the torus in order to flood the containment.
The licensee also identified several potential improvements for future use.
The most important are:
1.
Shedding the non-safety battery load so that it would be available to extend dc power supply after the safety batteries have failed during station blackout or using of a portable battery charger.
2.
Improved calibration of low vessel pressure emergency core cooling system permissive sensors.
3.
Capability to locally operate certain air-operated valves upon loss of instrument air.
4.
Increased drywell head preload to improve containment integrity at elevated temperatures.
5.
Modification of containment venting pressure in order to have high confidence that there is no large structural failure; and 6.
Improved operator training in areas where the IPE took credit for human error recovery.
Specifically, for recovering from loss of screenhouse
- intake, instrument air, and service water, and for controlling EC overfill events and operating the ECs after waterhammer events upon EC isolation.
3.0
~COIIC III aM On the basis of these findings, the NRC staff notes that:
(1) the licensee's IPE is complete with regard to the information requested by GL 88-20 (and associated guidance in NUREG-1335),
and (2) the IPE results are reasonable given the design, operation, and history of NMP1.
As a result, the staff concludes that the licensee's IPE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities and, therefore, the NMP1 IPE has met the intent of GL 88-20.
It should be noted that the staff's review primarily focused on the licensee's ability to examine NMP1 for severe accident vulnerabilities.
Although certain aspects of the IPE were explored in more detail than others, the review is not intended to validate the accuracy of the licensee's detailed findings (or
quantification estimates) that stemmed from the examination.
Therefore, this SER does not constitute NRC approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of GL 88-20.
Principal Contributors:
R. Clark E. Lois Date:
April 2, 1996
0 I
NINE MILE POINT NUCLEAR STATION UNIT NO. I INDIVIDUALPLANT EXAMINATION TECHNICAL EVALUATION REPORT (FRONT-END)
Enclosure 2
~
~