ML17157C363
| ML17157C363 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/09/1993 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | Byram R PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| TAC-M85668, TAC-M85669, NUDOCS 9306150220 | |
| Download: ML17157C363 (20) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 9, 1993 Docket Nos.
50-387 and 50-388 Mr. Robert G.
Byram Senior Vice President-Nuclear Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101
Dear Hr. Byram:
SUBJECT:
EMERGENCY ACTION LEVELS IN REVISED EMERGENCY PLAN, SUS(UEHANNA STEAM ELECTRIC STATION, UNITS 1
AND 2 (PLA-3893)
(TAC NOS.
M85668 AND M85669}
Your letter of January 21, 1993 forwarded a proposed revision to the.
Susquehanna Emergency Plan which implements new Emergency Action Levels (EALs) based on the Nuclear Management and Resources Council (NUHARC) NESP-007, Revision 2 methodology.
Your letter noted that you had taken some exceptions to the NUMARC/NESP-007 methodology and requested our approval of the revised plan.
We have completed our initial review of the proposed EALs.
Susquehanna is one of the lead boiling water reactor plants to incorporate the guidance in NUMARC/NESP-007, Revision 2, "Methodology for Development= of Emergency Action Levels."
NUHARC/NESP-007 was recently endorsed in Revision 3 to Regulatory Guide 1.101, "Emergency Planning and Preparedness for Nuclear Power Reactors,"
as an alternative means by which licensees can meet the requirements of 10 CFR 50.47(b}(4} and Section IV.B of Appendix E to Part 50.
Enclosed is our review of the Susquehanna EAI.s against the NUHARC/NESP-007 guidance.
Because of the staff's previous endorsement of this guidance, the review focused on those EALs that deviated from the guidance and those EALs that required the development of site-specific thresholds.
We have concluded that additional information is needed regarding a number of Susquehanna EALs which deviate from the NUMARC/NESP-007 guidance.
The details of the request for information are enclosed.
This requirement affects fewer than 10 respondents and, therefore, is not subject to Office of Management and Budget review under P.L.96-511.
930bl50220 930609 PDR
~ ] ADOCN 05000387
'DR~(
gRI; PI P I;PNTB C~Y
'I Mr. Robert G.
Byram June 9, 1993
\\
If you have any questions or wish to discuss our comments, please contact either the lead technical reviewer',
James B. O'rien at (301) 504-2919 or myself at (301) 504-1402.'
Sincerely,
Enclosure:
Review of EALs and Request for Additional Information cc w/enclosure:
See next page DISTRIBUTION:
Docket Fi1 e NRC
& Local PDRs PDI-2 Reading SVarga JCalvo CHiller HO'Brien(2)
RClark RErickson,9H15
, JO'Brien, 9H15
EWenzinger, RGN-I
- JWhite, RGN-I ar7gTnaT stgned by Richard J. Clark Richard J. Clark, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation OFFICE P
NAME DATE PDI-H
- tlc 93 Q4 g
93 PDI-2 D
CHI 93 OFFICIAL" R CORO COPY FILENAME: SUH85668.RAI
Hr. Robert G.
Byram.
June 9, 1993 If you have any questions or wish to discuss our comments, please contact either the lead technical
- reviewer, James B. O'rien at (301) 504-2919 or myself at (301) 504-1402.
Sincerely,
Enclosure:
Review of EALs and Request for Additional Information cc w/enclosure:
See next page J.
lark, Senior Project Nanager Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Mr. Robert G.
Byram Pennsylvania Power
& Light Company Susquehanna Steam Electric Station, Units 1
& 2 CC:
Jay Silberg, Esq.
- Shaw, Pittman, Potts
& Trowbridge 2300 N Street N.W.
Washington, D.C.
20037 Bryan A. Snapp, Esq.
. Assistant Corporate Counsel Pennsylvania Power
& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Mr. Harold G. Stanley Superintendent of Plant Susquehanna Steam Electric Station Pennsylvania Power and Light Company Box 467 Berwick, Pennsylvania 18603 Mr. Herbert D. Woodeshick pecial Office of,the President ennsylvania Power and Light Company Rural Route 1,
Box 1797 Berwick, Pennsylvania 18603 Mr. Scott Barber Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O.
Box 35 Berwick, Pennsylvania 18603-0035 George T. Jones Manager-Engineering Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Resources Commonwealth of Pennsylvania P. 0.
Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C. Tilton, III, Allegheny Elec. Cooperative, Inc.
212 Locust Street P.O.
Box 1266 Harrisburg, Pennsylvania 17108-1266 Mr. J.
M. Kenny Licensing Group Supervisor Pennsylvania Power
& Light Company S
2 North Ninth Street P
Allentown, Pennsylvania 18101
ENCLOSURE SUSQUEHANNA'TEAM-'ELECTRIC',STATIO'N;: ',,,::
<ANUARY::2'1'993"""""'MERGENCY" The January 21, 1993, proposed revision to Susquehanna Steam Electric Station (SSES) emergency action levels (EALs) was reviewed against the requirements in 10 CFR 50.47 (b)(4) and Appendix E to 10 CFR Part 50.
10 CFR 50.47(b)(4) specifies that onsite emergency plans must meet the following standard:
"A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee..."
Appendix E (IV)(C) specifies that "emergency action levels (based not only on onsite and offsite radiation monitoring information but also on readings from a number of sensors that indicate a potential emergency, such as the pressure in containment and the response of the Emergency Core Cooling System) for notification of offsite agencies shall be described.....
The emergency classes defined shall include: (1) notification of unusual events, (2) alert, (3) site area emergency, and (4) general emergency."
In Revision 3 to Regulatory Guide 1.101, "Emergency Planning and Preparedness for Nuclear Power Reactors,"
the NRC endorsed NUMARC/NESP-007, Revision 2 (NESP-007),
"Methodology for Development of Emergency Action Levels," as an acceptable method for licensees to meet the requirements of 10 CFR 50.47(b)(4) and Appendix E to 10 CFR Part 50.
The staff used NESP-007 as the basis for its review of the SSES proposed EAL revision.
The staff has identified a number of EALs which deviate from the NESP-007 guidance, which are listed below.
These EALs should be modified to meet the NESP-007 guidance or additional information must be provided which justifies how the EALs meet the intent of the NESP-007 guidance or otherwise meet the requirements in 10 CFR 50.47 and Appendix E to 10 CFR Part 50.
1.
general No Initiating Conditions (ICs)
The SSES EALs are not grouped under initiating conditions (ICs) as is specified in NESP-007.
Per NESP-007, ICs are "one of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency, or such an emergency has occurred."
EALs are "a pre-determined, site-specific, observable threshold for a plant initiating condition that places the plant in a given emergency class."
Although EALs are not required to be grouped under ICs by the regulations, the use of ICs is advantageous from a human factors perspective.
Grouping EALs under ICs willindicate to those who must use the EALs how an EAL (or several diverse EALs) is related to the plant condition which is of concern.
This willassist the emergency director in the use ofjudgement in making the correct event classifica-tion. The lack of ICs for loss of fission product barriers is of particular concern to the staff.
SSES should include ICs with their EALs where appropriate.
2.
EA 1]i 'h i*~d i
I<<
t iiO'echnical specification limits (Tech Spec 3.4.5 4 pCilgm dose equivalent l-131 This EAL does not include the technical specification coolant activity limitof 100/E-Bar as a condition and therefore deviates from NESP-007, IC SU4, EAL."
2.
This EAL should be revised to include the 100/E-BAR coolant activity limit or additional information should be provided to justify this deviation.
This EAL also deviates from NESP-007, IC SU4 in that Mode 5 is not included as an applicable mode.
This EAL should be revised to include Mode 5 as an applicable mode or additional information should be provided to justify this deviation.
3.
~EAL I i Reactor coolant activity determined by sample () 2000 ttCi/gm dose equi valent 1-131)
This EAL is classified as a Site Area Emergency.
This EAL deviates from NESP-007, IC FG1 "loss of any two barriers and potential loss of third barrier" which is classified as a General Emergency.
According to the basis for the SSES EAL, coolant activity at this level is indicative of 20% clad failure and is also indicative of a loss of the RCS barrier.
In the basis for the NESP-007 guidance for this EAL, it is stated that "regardless of whether containment is challenged, this amount of activity in containment, if
- released, could have such severe consequences that it is prudent to treat this as a potential loss of containment such that a General Emergency'eclaration is warranted."
'SSES EAL reference number and EAL description This EAL should be revised to correspond to the General Emergency classification level or additional information should be provided to justify this deviation.
4.
EAL 2 1.1 Visual observation ofa sustained, uncontrolled water level decrease in the reactor refueling cavity with all irradiated fuel assemblies remaining covered by water This EAL deviates from NESP-007, IC AA2, EAL 3 in that "defueled" is not included as an applicable mode.
This EAL should be revised to include the defueled mode as an applicable mode or additional information should be provided to justify this deviation.
I 5.
EAL214 2
p the top ofactive fuelfor ) 20 minutes This EAL appears to deviate from NESP-007, IC, FG1, EAL "Reactor vessel water level less than (site specific) value and the maximum core uncovery time limit is in the UNSAFE region."
From the supporting information provided with the SSES EAL it could not be determined whether the 20 minute time limit specified in the EAL corresponded to the NESP-007 EAL condition of the maximum core uncovery time limitbeing in the UNSAFE region.
Additional information should be provided regarding the relationship of the 20 minutes specified in the SSES EAL to the maximum core uncovery time limit being in the UNSAFE region specified in the NESP-007 EAL. Ifthe SSES EAL does not correspond to the NESP-007 EAL condition of the maximum core uncovery time limitbeing in the UNSAFE region, then the SSES EAL should be revised to be consistent with the NESP-007 EAL or addition information should be provided to justify the deviation.
6.
E~AL214 Il p
I I
I 5 d I
d reactor pressure vessel flooding pressure cannot be met within 20 minutes This EAL appears to deviate from NESP-007, IC FGl, EAL "Reactor vessel water level less than (site specific) value and the maximum core uncovery time limit is in the UNSAFE region."
From the supporting information provided with the SSES EAL, it could not be determined whether the 20-minute time limitspecified in the EAL corresponded to the NESP-007 EAL condition of the maximum core uncovery time limitbeing in the UNSAFE region.
Additional information should be provided regarding the relationship of the 20 minutes specified in the SSES EAL to the maximum core uncovery time limit being in the UNSAFE region specified in the NESP-007 EAL. Ifthe SSES EAL does not correspond to the maximum core uncovery time limitbeing in the UNSAFE region then the SSES EAL should be revised to be consistent with the NESP-007 EAL or addition information should be provided to justify the deviation.
7.
EAL 2 2 1
ttfullreactor scram has been initiated withfailure ofautomatic trips to bring the reactor subcritical This EAL deviates from NESP-007, IC SA2, "Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram was Successful," in that SSES classifies this event as an Unusual Event while the NESP-007 guidance classifies this as an Alert.
As stated in NESP-007, the failure of the RPS to automatically scram the reactor is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient.
Thus plant safety has been compromised and design limits of the fuel may have been exceeded.
An Alert is appropriate because conditions exist that lead to a potential loss of the fuel clad or reactor coolant system.
This EAL should be revised so that it is classified as an Alert or additional information should be provided to justify this deviation.
8.
EAA222 Afullreactor scram has been initiated withfailure ofboth automatic and manual trips to bring the reactor subcritical SSES classifies the failure of the automatic and manual reactor protection system as an Alert ifreactor power is less than 5 percent following initiation of the automatic and manual scram and as a Site Area Emergency ifpower is greater than 5 percent following the scram.
In NESP-007 IC SA2 and SS2, the failure of automatic RPS with a Zi~~fi3 manual scram is classified as an Alert and the failure of both the automatic and manual RPS scram is classified as an Site Area Emergency.
A successful manual scram is any set of actions by the operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical. Ifthe reactor is not subcritical following the manual scram, the event should be classified as a Site Area Emergency.
This EAL should be modified to include the condition that the manual scram was successful in bringing the reactor subcritical or additional information should be provided which justifies including a reactor power level after the scram instead of a successful manual scram.
This justification should provide the technical basis for the maximum power level specified in the SSES EAL as the delineation between the Alert and Site Area Emergency, including (1) the capability of safety systems to remove the heat produced from the specified reactor power level plus 100 percent decay heat (with containment isolation), (2) the expected operator actions for this type of an event as guided by the Emergency Operating Procedures, and (3) the plant condition and corresponding EAL which willresult in the escalation to the Site Area Emergency classification level.
- 9. E~Ll..b RPV I
l 9 d 9
d Mode 5 is not included as an applicable mode for this EAL: Although the NESP-007 does not specify applicable modes for this EAL, additional information is needed to determine whether or not mode 5 should be included as an applicable mode for the SSES EAL.
- 10. 5~2~2 Afullreactor scram has been initiated withfailure ofboth automatic and manual trips to bring the reactor subcritical, and reactor Power > 5 percent Afullreactor scram has been initiated withfailure ofboth automatic and manual trips to bring the reactor subcritical, and reactor power > 5 percent, and suppression pool temperature
> 200F This EAL deviates from NESP-007 IC SS2 in that the SSES EAL specifies that "Reactor Power is > 5 Percent" while the NESP-007 EAL specifies that "the manual scram was not successful."
For the reasons discussed under EAL 2.2.2 above, the condition of "Reactor Power > 5 Percent," should to be changed to "manual scram was not successful" for EAL 2.2.3 or additional technical justification for including the power level in the EAL should be provided.
- 11. EAL224 This EAL deviates from NESP-007 IC SG2 in that the SSES EAL specifies that "Reactor Power is > 5 Percent," while the NESP-007 EAL specifies that "the manual scram was not successful."
For the reasons discussed under EA'L 2.2.2 above, the condition of "Reactor Power > 5 Percent," should be changed to "manual scram was not successful" for EAL 2.2.4 or additional technical justification for including the power level in the EAL should be provided.
In addition, this EAL specified "Suppression Pool Temperature > 200'F" as the sole indication of the corresponding NESP-007 EAL conditions that "core cooling is extremely challenged" or "indication exists that heat removal is extremely challenged."
'Additional information should be provided which justifies not including other site specific indications for the corresponding NESP-007 EAL conditions.
12.
E~AL 1
1 Drywellpressure indication > 1.72 psig and indication ofa leak into containment This SSES EAL is classified as an Unusual Event.
This EAL deviates from NESP-007 IC FA1, which is classified as an Alert.
In the basis for the NESP-007 EAL it is stated that "the drywell pressure setpoint
~should be based on the drywell high pressure alarm set point and indicates a LOCA. A higher value may be used ifsupporting documentation is provided which indicates the chosen value is less than the pressure which would be reached for a 50 gpm RCS leak."
The documentation provided by SSES supporting this EAL indicates that 1.72 psig willbe reached ifthere is a 50 gpm leak. It is appropriate that an Alert be declared for a leak of this magnitude.
This EAL should be modified to correspond to the Alert level classification or additional information should be provided to justify this deviation.
13.
~EAL
.1 2 Drywell pressure indication > 3.0 psig and indication ofa leak into containment As discussed for EAL 3.1.1, NESP-007 states that "the drywell pressure setpoint for this EAL should be based on the drywell high pressure alarm set point and indicates a LOCA. A higher value may be used ifsupporting documentation is provided which indicates the chosen value is less than the pressure which would be reached for a 50 gpm RCS leak."
The documentation provided by SSES supporting this EAL indicates that 1.72 psig will be reached ifthere is a 50 gpm leak.
Therefore the setpoint for this EAL should be changed to 1.72 psig.
This EAL should be modified so that the "Drywell Pressure Indication" limitis changed to 1.72 psig or additional information should be provided to justify this deviation.
14.
~EAL 1
Rapid, unexplained decrease in drywel! pressure followinginidal pressure rise above 3.0 psig and indication ofa leak into containment As discussed for EAL 3.1.2, this EAL should be modified so that the "Drywell Pressure Indication" limitis changed to 1.72 psig or additional information should be provided to justify this deviation.
15.
~EAL 1 4 Drywellpressure and suppression chamber pressure exceed containment maximum internal pressure Events which result in Drywell gr suppression chamber pressures of this magnitude are indicative of a loss of RCS, potential loss of fuel clad and potential loss of containment.
NESP-007 does not contain an EAL for high drywell pressure which is indicative of a condition of potential loss of fuel clad.'his deviation is acceptable since the high drywell pressure is indicative of a major loss of RCS which could potentially damage the fuel clad.
Consistent with this conclusion, either high drywell pressure gr high suppression chamber pressure should result in the same classification.
Therefore, this EAL should be modified to indicate that either drywell pressure gr suppression chamber pressure exceeding containment maximum internal pressure will result in an Alert classification.
16.
~EAL
.2 1
Unidentified or pressure boundary leakage > 10 gpm This EAL deviates from NESP-007, IC SU5, EAL 2, "Identified Leakage Greater than 25 gpm," in that identified leakage is not included in this EAL.
The licensee's basis for not having an EAL for identified leakage is that the leakage comes from known, monitored locations such as valve packing and pump seals which do not represent a "leak before break" concern.
The NESP-007 IC for RCS'eakage specifically included an EAL for identified leakage.
In the basis for this IC it is stated that the RCS leakage IC is included as an Unusual Event because it may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of the safety of the plant.
This EAL should be modified to include the condition of "identified leakage greater than 25 gpm" or additional justification for not including identified leakage should be provided,
- 17. EAL 4 1 2 rin unisolable primary system leak is discharging inro secondary containment and secondary containment temperature exceeds maximum safe temperature limitin one area of Table 4.1.1 This EAL deviates from one of the example EALs for NESP-007, IC FS1, i.e.
"unisolable primary sys leakage outside drywell as indicated by area temp or area rad alarm."
The NESP-007 EAL is indicative of a loss of the RCS and Primary Containment fission product barriers and the condition is classified as a Site Area Emergency.
In contrast to this, EAL 4.1,2. is classified as an Alert.
'n the basis for this EAL the licensee states that this EAL represents a challenge to both the RCS and containment fission product barriers.
EAL 4.1.3, "An Unisolable Primary System Leak is discharging into Secondary Containment and Secondary Containment Temperature exceeds Maximum Safe Temperature Limit in MORE than one area of Table 4.1.1," is classified as a Site Area Emergency=;".
and in the basis it is stated that this EAL is indicative of a potential loss of RCS;-.
and loss of containment fission product barriers.
.The licensee uses the indication of more than one area exceeding the safe temperature limitversus one area exceeding the safe temperature limitas a discriminator on the size of the leak, This EAL should be revised to correspond to the Site Area Emergency classification level or information should be provided as to the size of the leak which would result in one area temperature exceeding the maximum safe area temperature and the size of the leak which would result in more than one area and to further justify this deviation from the NESP-007 guidance.
18.
~EAL 4 1.
An unisolable primary system leak is discharging into secondary containment and secondary containment temperature exceeds maximum safe temperature limitin MORE than one area of Table 4.1.1 For the reasons discussed under EAL 4.1.2, this EAL should be revised to change the condition that "MORE than one area...." to "any area" or information should be provided as to the size of the leak which would result in one area temperature exceeding the maximum safe area temperature and the size of the leak which would result in more than one area exceeding the safe temperature limitand to further justify this deviation from the NESP-007 guidance.
- 19. EAL414 An unisolable primary system leak is discharging into secondary containment and secondary containment temperature exceeds maximum safe temperature limitin MORE than one area of Table 4.1.1 and any parameter in Table 4.1.2, "Indication of Fuel Cladding Degradation," has been exceeded The condition that the "Maximum Safe Temperature Limitis exceeded" is unnecessary in this EAL, since there is sufficient indication that all three fission product barriers have been lost without including this condition.
The condition that the Maximum Safe Temperature Limitis exceeded should be deleted or additional information which justifies including this condition should be provided.
20.
~EAL 1
Dose calculations are unavailable and nvo consecutive 10 minute average SPING vent monitor readings indicate:
Noble
-.'as release rate > 2.7x 10',Cilmin or Iodine-131 release rate
> 5.3x10',Cilmin In EALs 5.1.1 and 5.1.2, the particulate release rate is included as an indication of radioactive releases.
Additional information should be provided which justifies not including the condition that, the Particulate Release rate is above a (site specific) set point, as part of this EAL.
21.
~EAL 2
Actual or projected dose at or beyond the EPB whole body
>100 mrem or child thyroid > 500 mrem An unisolable primary system leak is discharging into secondary containment and secondary containment radiation levels exceed maximum safe radiation level in one area ofTable $.441 The definition of the EPB (emergency planning boundary) was not specified in the EAL basis.
Additional information should be provided concerning the definition of the EPB to determine whether this EAL meets NESP-007 IC AS1.
"unisolable primary sys leakage outside drywell as indicated by area temp or area rad alarm."
The NESP-007 EAL is indicative of a loss of the RCS and Primary Containment fission product barriers and the condition is classified as a Site Area Emergency.
In contrast to this, EAL 5.4.2.c is classified as an Alert.
In the basis for EAL 5.4.2.c, it is stated that this EAL represents a challenge to both the RCS and containment fission product barriers.
EAL 5.4.3, "An Unisolable Primary System Leak is discharging into Secondary Containment and Secondary Containment Radiation level exceed Maximum Safe Radiation level in MORE than one area of Table 5.4.1," is classified as a Site Area Emergency and in the basis it is stated that this EAL is indicative of a potential loss of RCS and loss of containment fission product barriers.
The licensee uses the indication of more than one area exceeding the safe radiation level versus one area exceeding the safe radiation level as a discriminator on the size of the leak.
This EAL should be revised to change the condition that "MORE than one area..." to " any area..." or information should be provided as to the size of the leak which would result in the one area radiation level exceeding the maximum safe radiation level and the size of the leak which would result in more than one area exceeding the maximum safe radiation level and to further justify this deviation from the NESP-007 guidance.
23.
~EAL
.4.
An unisolable primary system leak is discharging into secondary containment and secondary containment radiation level exceeds maximum safe radiation level in MORE than one area of Table 5.4. I For the reasons discussed in EAL 5.4.2.c, this EAL should be revised to change the condition that "MORE than one area...." to "any area" or information should be provided as to the size of the leak which would result in the one area radiation level exc'ceding the maximum safe area radiation level and the size of the leak which would result in more than one area exceeding the maximum safe radiation level and to further justify this deviation from the NESP-007 guidance.
- 24. ~EL ~44 An unisolable primary system leak is discharging into secondary containment and secondary containment radiation level exceeds maximum safe radiation level in MORE than one area of Table 5.4.1 and any parameter in Table 5.4.2, "Indication ofFuel Cladding Degradation, " has been exceeded The condition that the Maximum Safe Radiation level is exceeded is unnecessary in this EAL, since there is sufficient indication that all three fission product barriers have been lost without including this condition.
The condition that the Maximum Safe Radiation level is exceeded should be deleted or additional information which justifies including this condition should be provided.
25.
~EAL i
Loss ofpower Pom stan up-transformers 10 and 20 to either unit and ALL4.15 KVESS buses on either unit are de-energized
> 15 minutes In the basis for NESP-007 IC SS1, it is stated that time duration for the loss of power should be selected to exclude transient or momentary power losses, but should not exceed 15 minutes.
In the basis, for EAL 6,1.3 it is stated that 15 minutes has been selected to exclude transient or momentary power losses.
Further information should be provided to justify why 15 minutes is appropriate for SSES.
- 26. EAL7 2
Uncontrolled reactor coolant temperature increase (>200'F),
~
and Isuppression pool temperature > 120'F or suppression pool evel ~ 12fij This EAL deviates from NESP-007, IC SA3, "inability to maintain plant in cold shutdown."
One of the EALs for this NESP-007 IC is; "Loss of (site-specific Technical Specification required function to maintain cold shutdown."
SSES EAL 7.3.2.a specifies the condition of high suppression pool temperature and low suppression pool level as indications for the loss of Technical Specification required functions.
These specific conditions may not be the only indications or earliest site specific indications of this condition.
This EAL should be modified to include additional indications, such as loss of RHR, as may be appropriate for SSES.
27.
Confirmed security threat directed toward the station This EAL deviates from NESP-007, IC HU4, in that site specific safeguard events are not delineated in the EAL, i.e. the SSES EAL does not indicate what constitutes a "security threat."
The basis for this EAL contains a list of examples of security threats, however, it appears that the individual classifying the event willnot be able to properly classify the event without referring to the basis.
This EAL should be modified to include site specific indications of "security. threats" or additional information should be provide to justify this deviation from NESP-007.
28.
~EAL 1 2 Confirmed hostile intrusion or aet in the plant protested area The corresponding NESP-007 IC to this EAL is HA4, which contains the following as EALs:
1.
Intrusion into plant protected area by a hostile force 2.
Other security events as determined from (site-specific) safeguards Contingency Plan.
Additional information is needed to justify that "hostile intrusion or act" is consistent with the NESP-007 guidance for "other security events" and to justify how the SSES EAL is related to the SSES Safeguards Contingency Plan.
29.
~EAL i
Confirmed hostile intrusion or act in the plant vital areas The corresponding NESP-007 IC to this EAL is HS1, which contains the following as EALs:
1.
Intrusion into plant vital area by a hostile force 2.
Other security events as determined from (site-specific) safeguards Contingency Plan.
Additional information is needed to justify that "hostile intrusion or act" is consistent with the NESP-007 guidance for "other security events" and to justify how the SSES EAL is related to the SSES Safeguards Contingency Plan.
- 30. ~EL 22 R
llyly S
I d
d 2 f yl safety systems required for the current operating mode The corresponding NESP-007 is IC HA2, "Fire or Explosion Affecting the Operability of plant safety systems required to establish or maintain safe shutdown."
One of the EALs for NESP-007, IC HA2, is as follows:
1.
The following conditions exist:
a.
Fire or explosion in any of the following areas (site specific) and b.
Affected system parameter indications show degraded performance or plant personnel report visible damage to permanent structures or equipment with the specified area The SSES EAL deviates from the NESP-007 guidance in that no site specific "areas" are listed and in that the SSES EAL specifies that the fire affects redundant trains, The intent of the NESP-007 guidance for this IC is that only one of the redundant trains needs to be affected to declare an Alert. The concern is with the magnitude of the fire and damage to a single train of a safety system is used as an indication of a fire of sufficient magnitude to warrant an Alert to be declared.
This EAL should be modified to include a condition which indicates the "specific areas" affected by the fire or additional information should be provided which '-',
justifies not including specific areas.
In addition this EAL should be modifiei to change the EAL condition of "Fire likely to affect the redundant trains" to "Fire, likely to affect a train." The licensee may propose a different condition for this~~
EAL which willbe indicative of a fire of sufficient magnitude to warrant an Alert declaration consistent with the basis for this NESP-007 EAL.
- 31. ~EL 2.2.3 Report or detection oftoxic orflammable gases w'ithin a plant vital structure in concentration that willbe life threatening to plant personnel or willaffect plant systems required to establish or maintain cold shutdown This EAL deviates from NESP-007, IC HA3, "Release of Toxic or Flammable Gases within a Facility Structure..." in that the SSES EAL specifies plant "Vital Structure" while the NESP-007 specifies "Facility Structure."
The basis for the NESP-007 EAL states that "this IC applies to building and areas contiguous to the plant.Vital Areas..."
This EAL should be modified to include areas contiguous to the plant Vital areas or additional information should be provided to justify this deviation from NESP-007.
32.
P-7 I 3
yl 2
2 fg 2'Egg &i yyg'o the environment that exceed two times the radiological technical specifications for 60 minutes or longer SSES did not include EALs corresponding to two of the EALs included under this NESP-007 IC.
'NESP-007, initiating condition (IC) reference number and IC description One of these EALs is related to readings on perimeter radiation monitoring systems.
SSES states that they do not have telemetered perimeter monitors.
A compensatory indication such as field monitoring readings should be used to provide the readings which are referred to in the NESP-007 EAL. SSES should include this type of an EAL or should provide additional justification for its omission.
The other EAL not included in the SSES submittal is an EAL related to the automatic real-time dose assessment capability.
SSES states that they do not use automatic initiation of real time dose assessment.
A compensatory indication such as manual initiation of real time dose assessment should be included in their EALs.
SSES should include this type of an EAL or should provide additional justification for not including this type of an EAL. This comment also applies to NESP-007, IC AA1.
- 33. ~yl I
yy yp <<I II f SSES does not include an EAL for loss of containment which corresponds to the
.NESP-007 IC. The licensee should provide additional information which justifies not including an Unusual Event level EAL for the loss or potential loss of containment.
In addition SSES does not include a potential containment loss EAL which corresponds to the NESP-007 EAL "explosive mixture exists."
SSES states "for an explosive mixture to exist a LOCA with Fuel Damage must have occurred and that these conditions are covered by the Drywell Rad Monitor and the Reactor Water Level thresholds."
The presence of an explosive mixture in the drywell is a valid indication of the potential loss of primary containment barrier.
Barrier based EALs are not
.- derived from a given accident sequence, rather they reference available indications that a barrier is potentially lost or lost.
3 SSES should add this EAL or provide additional justification for this deviation from the NESP-007 guidance.
- 34. ~P-7 7 ll
- fl DR ICld dRCS SSES did not tie the reactor coolant activity (indication of a loss of Fuel clad) with the Drywell Pressure (indication of a loss of RCS) as being indicative of a loss of two barriers and therefore being classified as a Site Area Emergency per the NESP-007 guidance.
SSES states that they elected not to explicitly tie Reactor Coolant Activityand Drywell Pressure because of the extended time required to obtain a reactor coolant sample.
In addition, they stated that this combination is considered unnecessary because other indicators cover this event with more timely indications.
Elevated reactor coolant activity is a valid indication of the loss of Fuel Clad Barrier.
Barrier-based EALs are not derived from a given accident sequence, rather they reference available indications that a barrier is potentially lost or lost.
It is appropriate and in accordance with NESP-007 to include this barrier-based EAL.
SSES should add this EAL or provide additional justification for this deviation
~from the NESP-007 guidance.
35.
SSRS
-DD7 Gl I.
ffd I l&,S <<I II fRCS&I*f, containment NESP-007 guidance contains as an initiating condition (FGl) for the General Emergency class, the loss any two barriers with the potential loss of a third barrier.
SSES did not tie the reactor coolant activity (indication of a loss of Fuel clad) with the RCS leakage (indication of a potential loss of RCS) and with indications for loss of Containment and therefore does not have in this case an EAL which corresponds to the NESP-007 guidance.
SSES states that they elected not to explicitly tie Reactor Coolant Activityand Drywell Pressure because of the extended time required to obtain a reactor coolant sample.. In addition, they stated that this combination is considered unnecessary because other indicators cover this event with more timely indications.
Elevated reactor coolant activity is a valid indication of the loss of Fuel Clad Barrier.
Barrier-based EALs are not derived from a given accident sequence; rather, they reference available indications that a barrier is potentially lost or lost. It is appropriatecand in accordance with NESP-007 to include this barrier-based EAL.
SSES should add this EAL or provide additional justification for this deviation from the NESP-007 guidance,
- 36. ~P-7 3
\\
3 I Ad*<<i pp f Ig 3 protected area EAL 7 for this NESP-007 IC is other Site-sp'ecific Occurrences.
Examples given in the basis for this EAL includes floods.
SSES did not include this in their EALs.
Internal flooding of the plant should be added as an EAL or addition information justifying not including this EAL should be provided.
37.~d pl*
I ff<<l dd II I
I 3
shutdown SSES does not have an EAL which corresponds to NESP-007 IC SS4.
SSES states that the NESP-007 EAL is applicable to PWRs and that there is no analogous situation for a BWR.
.The staff agrees that the NESP-007 EAL is tailored for PWRs.
However the staff believes that a similar EAL for BWR should be developed for complete loss of function needed to achieve cold shutdown.