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{{#Wiki_filter:r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board I'                             .
{{#Wiki_filter:r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board I'
In the Matter of                             )
In the Matter of
                                                        )
)
LONG ISLAND LIGHTING COMPANY                 )   Docket No. 50-322 (OL)
)
                                                        )
LONG ISLAND LIGHTING COMPANY
(Shoreham Nuclear Power Station,               )
)
Unit 1)                                     )
Docket No. 50-322 (OL)
AFFIDAVIT OF DR. GLENN G.       SHERWOOD, DR. ATAMBIR S. RAO, AND MR. EUGENE C.                   ECKERT Glenn G. Sherwood, Atambir S. Rao, and Eugene C. Eckert being duly sworn, depose and state as follows:
)
(1)   My name is Glenn G. Sherwood.                     I am employed by the General Electric Company as Manager, Safety and Licensing Operation.     My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.                         I have been employed in this position since 1976.           My responsibilities in-clude supervision of the preparation of licensing submittals for General Electric BWRs, including analyses performed in Chapter 15 of safety analysis reports.           In particular, I have been involved in the supervision of licensing matters for the Shoreham Nuclear Power Station since the initial submittal of the Shoreham Final Safety Analysis Report (FSAR).                         In this re-gard, I am familiar with the analyses performed in Chapter 15' l
(Shoreham Nuclear Power Station,
of that document.           From 1974, when I joined General Electric,                               I 8404020252 840331 PDR ADOCK 05000322                                                                                   ,
)
O                  PDR       a;                                                                   ,1 l
Unit 1)
)
AFFIDAVIT OF DR. GLENN G.
: SHERWOOD, DR. ATAMBIR S.
RAO, AND MR. EUGENE C.
ECKERT Glenn G.
Sherwood, Atambir S. Rao, and Eugene C. Eckert being duly sworn, depose and state as follows:
(1)
My name is Glenn G.
Sherwood.
I am employed by the General Electric Company as Manager, Safety and Licensing Operation.
My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.
I have been employed in this position since 1976.
My responsibilities in-clude supervision of the preparation of licensing submittals for General Electric BWRs, including analyses performed in Chapter 15 of safety analysis reports.
In particular, I have been involved in the supervision of licensing matters for the Shoreham Nuclear Power Station since the initial submittal of the Shoreham Final Safety Analysis Report (FSAR).
In this re-gard, I am familiar with the analyses performed in Chapter 15' of that document.
From 1974, when I joined General Electric, 8404020252 840331 PDR ADOCK 05000322 O
PDR a;
,1


l l
> to 1976, I was the Manager, Program Control Section.
to 1976, I was the Manager, Program Control Section. My responsibilities included managing engineering and i             .
My responsibilities included managing engineering and i
manufacturing work flow for General Electric's nuclear group.
manufacturing work flow for General Electric's nuclear group.
I have a Bachelor of Science degree in Engineering from the U.S. Naval Academy and a Ph.D. in Engineering from the Univer-sity of Michigan.
I have a Bachelor of Science degree in Engineering from the U.S. Naval Academy and a Ph.D. in Engineering from the Univer-sity of Michigan.
,              (2)   My name is Atambir S. Rao. I am employed by the General Electric Company as Manager, Plant Safety Systems Engi-neering. My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125. I was appointed to my present position in 1984. My responsibilities include ECCS performance analysis, containment performance response analy-sis, and plant safety performance evaluations, including FSAR safety analyses. I have previously held a number of positions
(2)
;    relating to accident and transient analyses since I first 1
My name is Atambir S. Rao.
joined General Electric in 1973. Earlier responsibilities have included modeling and analyzing the thermal hydraulic _ behavior of BWR fuel following loss of coolant accidents, assessing the implication of advances in heat transfer, fluid mechanics, thermodynamics and two-phase flow on overall BWR system re-sponse during transients and loss of coolant accidents, devel-l
I am employed by the General Electric Company as Manager, Plant Safety Systems Engi-neering.
!    oping emergency operator guidelines, and assessing containment thermal hydraulic and radiological response for various I
My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.
I was appointed to my present position in 1984.
My responsibilities include ECCS performance analysis, containment performance response analy-sis, and plant safety performance evaluations, including FSAR safety analyses.
I have previously held a number of positions relating to accident and transient analyses since I first 1
joined General Electric in 1973.
Earlier responsibilities have included modeling and analyzing the thermal hydraulic _ behavior of BWR fuel following loss of coolant accidents, assessing the implication of advances in heat transfer, fluid mechanics, thermodynamics and two-phase flow on overall BWR system re-sponse during transients and loss of coolant accidents, devel-l oping emergency operator guidelines, and assessing containment thermal hydraulic and radiological response for various I


r-o
r-o
                ~
~ accidents and transients.
accidents and transients.       I have been assigned as Manager, O
I have been assigned as Manager, O
Emergency Core Cooling Systems (ECCS) Engineering (1979-80),
Emergency Core Cooling Systems (ECCS) Engineering (1979-80),
and Manager, Containment and Radiological Engineering (198"-84).-
and Manager, Containment and Radiological Engineering (198"-84).- I received a Ph.D and a Masters degree in Mechani-2 cal Engineering from the University of California, Berkeley, and a Bachelor of Technology in Mechanical Engineering from the Indian Institute of Technology, Kanpur, India.
2        I received a Ph.D and a Masters degree in Mechani-cal Engineering from the University of California, Berkeley, and a Bachelor of Technology in Mechanical Engineering from the Indian Institute of Technology, Kanpur, India.
(3)
(3)     My name is Eugene C. Eckert. I am employed by the General Electric Company as Manager, Power Transient Per-forming Engineering, a position I have held since 1971.         My business address is General Electric Company, 175 Curtner Ave-nue, San Jose, California 95125.         I am responsible for estab-lishing the simulation requirements of the computer models needed to perform transient analyses, development of design procedures evaluation of BWR stability, and evaluation and specification of the functional protection systems required for reactor abnormal transient protection.       Immediately upon joining General Electric Company in September 1959, I partici-pated in assignments which included large jet engine control design, aircraft nuclear propulsion control analysis, nuclear submarine kinetics and control analysis, and industrial control simulation analysis at GE's Research and Development Center.
My name is Eugene C.
Eckert.
I am employed by the General Electric Company as Manager, Power Transient Per-forming Engineering, a position I have held since 1971.
My business address is General Electric Company, 175 Curtner Ave-nue, San Jose, California 95125.
I am responsible for estab-lishing the simulation requirements of the computer models needed to perform transient analyses, development of design procedures evaluation of BWR stability, and evaluation and specification of the functional protection systems required for reactor abnormal transient protection.
Immediately upon joining General Electric Company in September 1959, I partici-pated in assignments which included large jet engine control design, aircraft nuclear propulsion control analysis, nuclear submarine kinetics and control analysis, and industrial control simulation analysis at GE's Research and Development Center.
In 1962, I joined General Electric's Nuclear Energy Division to
In 1962, I joined General Electric's Nuclear Energy Division to


i work on Boiling Water Reactor simulation and dynamic analysis.
> i work on Boiling Water Reactor simulation and dynamic analysis.
I have been responsible for design and licensing documentation of the 5 namic analysis for several GE BWRs and have partici-pated in initial startup testing of many of the units. I re-ceived a Bachelor of Science Degree in Electrical Engineering from Valparaiso University in Indiana in 1958. I attended Stanford University under an Oak Ridge Fellowship and received a Master of Science Degree in Engineering Science in August 1959.
I have been responsible for design and licensing documentation of the 5 namic analysis for several GE BWRs and have partici-pated in initial startup testing of many of the units.
(4)   Chapter 15 of the Shoreham FSAR provides the re-sults of analyses for the spectrum of accident and transient events that must be accommodated by the Shoreham plant to dem-onstrate compliance with the NRC's regulations. This portion of the safety analysis is performed to evaluate the ability of the plant to operate without undue risk to the health and safe-ty of the public. The Shoreham FSAR was submitted to the NRC Staff for review and has been approved by the Staff in its Safety Evaluation Report for Shoreham (NUREG-0420).
I re-ceived a Bachelor of Science Degree in Electrical Engineering from Valparaiso University in Indiana in 1958.
(5)   At the request of the Long Island Lighting Compa-ny, General Electric, in conjunction with cognizant LILCO and Stone & Webster personnel, has reviewed all of the events con-sidered in Chapter 15 of the FSAR to determine the effect on-public health and safety of the operation of the Shoreham plant i
I attended Stanford University under an Oak Ridge Fellowship and received a Master of Science Degree in Engineering Science in August 1959.
(4)
Chapter 15 of the Shoreham FSAR provides the re-sults of analyses for the spectrum of accident and transient events that must be accommodated by the Shoreham plant to dem-onstrate compliance with the NRC's regulations.
This portion of the safety analysis is performed to evaluate the ability of the plant to operate without undue risk to the health and safe-ty of the public.
The Shoreham FSAR was submitted to the NRC Staff for review and has been approved by the Staff in its Safety Evaluation Report for Shoreham (NUREG-0420).
(5)
At the request of the Long Island Lighting Compa-ny, General Electric, in conjunction with cognizant LILCO and Stone & Webster personnel, has reviewed all of the events con-sidered in Chapter 15 of the FSAR to determine the effect on-public health and safety of the operation of the Shoreham plant i


1 a
, a during fuel load, criticality testing and low power operations.
during fuel load, criticality testing and low power operations.
Although the FSAR considers all phases of the operation of the plant f$om fuel load to operation at 100% power, this review was performed specifically to confirm that operation of the Shoraham plant during low power operation will pose no undue risk to public health and safety.
Although the FSAR considers all phases of the operation of the plant f$om fuel load to operation at 100% power, this review was performed specifically to confirm that operation of the Shoraham plant during low power operation will pose no undue risk to public health and safety. The review of Chapter 15 was divided into three parts: (1) fuel load and precriticality testing (Phase I), (2) cold criticality testing (Phase II), and (3) low power testing up to 5% of rated power (Phases III and IV).1/   The review was based upon the same criteria and bases as the original Chapter 15 analyses. Where assumption of a loss or unavailability of offsite power was required in the original analyses, potential unavailability of the TDI diesel generators was considered in this review.
The review of Chapter 15 was divided into three parts: (1) fuel load and precriticality testing (Phase I), (2) cold criticality testing (Phase II), and (3) low power testing up to 5% of rated power (Phases III and IV).1/
(6)   The General Electric review of Chapter 15 con-firms that operation during the phases identified above will not result in any undue risk to the public health and safety.
The review was based upon the same criteria and bases as the original Chapter 15 analyses.
In fact, the risk from any Chapter 15 event during both the fuel load and precriticality phase and the cold criticality testing phase is essentially non-existent. The risk to the l
Where assumption of a loss or unavailability of offsite power was required in the original analyses, potential unavailability of the TDI diesel generators was considered in this review.
1/ Parts (1) and (2) correspond to Phases I and II, respec-tively, as described in the Affidavit of Messrs. Notaro and Gunther. Part (3) corresponds to Phases III and IV, combined, as described in that affidavit.
(6)
The General Electric review of Chapter 15 con-firms that operation during the phases identified above will not result in any undue risk to the public health and safety.
In fact, the risk from any Chapter 15 event during both the fuel load and precriticality phase and the cold criticality l
testing phase is essentially non-existent.
The risk to the 1/
Parts (1) and (2) correspond to Phases I and II, respec-tively, as described in the Affidavit of Messrs. Notaro and Gunther.
Part (3) corresponds to Phases III and IV, combined, as described in that affidavit.
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public health and safety from the Chapter 15 avents postulated for low power testing up to 5% of rated power is small in com-parison'to the risks already found acceptable for 100% power                                       -
- public health and safety from the Chapter 15 avents postulated for low power testing up to 5% of rated power is small in com-parison'to the risks already found acceptable for 100% power operation.' As already indicated, this review considered the impact of potential diesel unavailability.
operation.' As already indicated, this review considered the impact of potential diesel unavailability.
Phase I:
Phase I:                         Fuel Loading and Precriticality Testing (7)       This phase of operation of the Shoreham plant in-cludes only initial fuel loading and precriticality testing.
Fuel Loading and Precriticality Testing (7)
The reactor will remain at essentially ambient temperature and atmospheric pressure.                         The reactor will not be taken critical.
This phase of operation of the Shoreham plant in-cludes only initial fuel loading and precriticality testing.
Any increase in temperature beyond ambient conditions will be due only to external heat sources such'as recirculation pump heat. There will be no heat generation in the core.                                     Details of the steps to be performed during these operations are de-scribed in the Phase I discussion in the affide At submitted by Messrs. Notaro and Gunther.
The reactor will remain at essentially ambient temperature and atmospheric pressure.
(8) The review of the Chapter 15 analysis revealed that of the 38 accident or transient events addressed in Chap-ter 15, 18 of the events could not occur during Phase I because of the operating conditions of the plant.                                     An additional 5 events could physically occur, but given the plant conditions, could rat constitute events in the context of the Chapter 15
The reactor will not be taken critical.
Any increase in temperature beyond ambient conditions will be due only to external heat sources such'as recirculation pump heat.
There will be no heat generation in the core.
Details of the steps to be performed during these operations are de-scribed in the Phase I discussion in the affide At submitted by Messrs. Notaro and Gunther.
(8)
The review of the Chapter 15 analysis revealed that of the 38 accident or transient events addressed in Chap-ter 15, 18 of the events could not occur during Phase I because of the operating conditions of the plant.
An additional 5 events could physically occur, but given the plant conditions, could rat constitute events in the context of the Chapter 15


a   l J
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i          safety analysis.                         The remaining 15 events could possibly occur, although occurrence is highly unlikely given the plant condi-
J i
                                                                                                                                                ^
safety analysis.
tions.           In any event, it is readily apparent that the potential consequences of these 15 events would be trivial.                                                   Exhibit 1
The remaining 15 events could possibly occur, although occurrence is highly unlikely given the plant condi-
;          below lists the category into which each Chapter 15 event j         falls.
^
i (9)       The 18 Chapter 15 events which could not occur j         during Phase I are precluded by the operating conditions of the                                                                         ,
tions.
1                                                                                                                                                  l 1
In any event, it is readily apparent that the potential consequences of these 15 events would be trivial.
reactor.                 These events all involve operating modes or component operation which are not possible during this phase.                                                   For exam-i l         ple, during fuel loading and precriticality testing, the reac-1 i        tor is at essentially ambient temperature and atmospheric pres-sure.       Accordingly, no steam is available.                                 Thus, all events
Exhibit 1 below lists the category into which each Chapter 15 event j
!          which would require pressurized conditions are precluded.
falls.
Events such as turbine trip (FSAR 5 15A.1.2), loss of feedwater heating (FSAR 5 15A.1.8) and inadvertent opening of a safety j         relief valve require the generation of steam for the event to occur.           Similarly, there is no steam flow to interrupt, thus i          precluding an MSIV closure event (FSAR 5 15A.1.4).                                                   Other 1
(9)
events are precluded by definition.                                       Thus, events such as con-tinuous control rod withdrawal during power range operation (FSAR 5 15A.1.11) and operation of a fuel assembly in an.im-i proper location (FSAR 5 15A.1.16) cannot be. postulated.
The 18 Chapter 15 events which could not occur i
j during Phase I are precluded by the operating conditions of the 1
reactor.
These events all involve operating modes or component 1
operation which are not possible during this phase.
For exam-i l
ple, during fuel loading and precriticality testing, the reac-1 tor is at essentially ambient temperature and atmospheric pres-i sure.
Accordingly, no steam is available.
Thus, all events which would require pressurized conditions are precluded.
Events such as turbine trip (FSAR 5 15A.1.2), loss of feedwater heating (FSAR 5 15A.1.8) and inadvertent opening of a safety j
relief valve require the generation of steam for the event to occur.
Similarly, there is no steam flow to interrupt, thus precluding an MSIV closure event (FSAR 5 15A.1.4).
Other i
1 events are precluded by definition.
Thus, events such as con-tinuous control rod withdrawal during power range operation (FSAR 5 15A.1.11) and operation of a fuel assembly in an.im-i proper location (FSAR 5 15A.1.16) cannot be. postulated.
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                                        . _ . - ~ ,                 - . - _ _ . - _ _ .      _- _ - _ _ ,                . . . .
. _. - ~,


i (10) In addition to the 18 events which nimply cannot occur, there are 5 events for which the component operation evaluated in Chapter 15 could occur, but the phenomena of in-                   -
, i (10)
terest in Chapter 15 could not exist. All recirculation pump events, such as recirculation pump trip (FSAR $ 15A.1.20) and abnormal startup of an idle recirculation pump (FSAR 5 15A.1.25), would be of interest only if they could affect core physics or thermal-hydraulic conditions.       With no heat generation or boiling in the core, there are no pertinent phe-nomena (such as temperature differences or void collapses) to evaluate. Another example, the core coolant temperature in-crease event (FSAR $ 15.A.1.26), postulates a loss of RHR cool-ing. Even if the RHR system was operated in Phase I, there would be no temperature increase from decay heat to evaluate should the RHR system be lost.
In addition to the 18 events which nimply cannot occur, there are 5 events for which the component operation evaluated in Chapter 15 could occur, but the phenomena of in-terest in Chapter 15 could not exist.
(11) The remaining 15 events addressed in Chapter 15 could possibly occur. However, our review established that all are trivial events which have no potential to impact public health and safety. Prior to initial criticality, there are no fission products in the core and no decay heat exists.           It fol-lows that core cooling is not required. In addition, with no fission product inventory, there are no fission product re-leases possible. Thus, for reactor events such as a control
All recirculation pump events, such as recirculation pump trip (FSAR $ 15A.1.20) and abnormal startup of an idle recirculation pump (FSAR 5 15A.1.25), would be of interest only if they could affect core physics or thermal-hydraulic conditions.
With no heat generation or boiling in the core, there are no pertinent phe-nomena (such as temperature differences or void collapses) to evaluate.
Another example, the core coolant temperature in-crease event (FSAR $ 15.A.1.26), postulates a loss of RHR cool-ing.
Even if the RHR system was operated in Phase I, there would be no temperature increase from decay heat to evaluate should the RHR system be lost.
(11)
The remaining 15 events addressed in Chapter 15 could possibly occur.
However, our review established that all are trivial events which have no potential to impact public health and safety.
Prior to initial criticality, there are no fission products in the core and no decay heat exists.
It fol-lows that core cooling is not required.
In addition, with no fission product inventory, there are no fission product re-leases possible.
Thus, for reactor events such as a control


    .                                                                    l
, ]
  ,                                                                     1
rod removal error (FSAR 5 15A.1.13) and a control rod drop (FSAR $ 15.1.33) and for non-reactor events such as a fuel han-dling abcident (FSAR $ 15.1.36) or a liquid radwaste tank rup-ture (FSAR'l 15.1.32), there could be no radiological conse-quences.
                                                                                                            ]
Therefore, there is no risk to public health and safety.
rod removal error (FSAR 5 15A.1.13) and a control rod drop         1 (FSAR $ 15.1.33) and for non-reactor events such as a fuel han-dling abcident (FSAR $ 15.1.36) or a liquid radwaste tank rup-ture (FSAR'l 15.1.32), there could be no radiological conse-quences. Therefore, there is no risk to public health and safety.
(12)
(12) Even a loss of coolant accident (FSAR $ 15.1.34) could have no radiological consequences during Phase I. No core cooling is required. No fission product release is possi-ble. The fuel simply could not be challenged by a complete draindown of the reactor vessel for an unlimited period of time.
Even a loss of coolant accident (FSAR $ 15.1.34) could have no radiological consequences during Phase I.
(13) In summary, the review of Chapter 15 events for fuel loading and precriticality testing indicates that nany Chapter 15 events simply cannot occur, and for those that can, there can be no radiological consequences. Therefore, there is no possible risk to the public health and safety. This conclu-sion is not affected by any postulated diesel generator unavailability because it is in no way dependent on the avail-ability or unavailability of any AC power.
No core cooling is required.
Phase II: Cold Criticality Testing (14) This phase of low power testing of the Shoreham
No fission product release is possi-ble.
The fuel simply could not be challenged by a complete draindown of the reactor vessel for an unlimited period of time.
(13)
In summary, the review of Chapter 15 events for fuel loading and precriticality testing indicates that nany Chapter 15 events simply cannot occur, and for those that can, there can be no radiological consequences.
Therefore, there is no possible risk to the public health and safety.
This conclu-sion is not affected by any postulated diesel generator unavailability because it is in no way dependent on the avail-ability or unavailability of any AC power.
Phase II:
Cold Criticality Testing (14)
This phase of low power testing of the Shoreham


      -                                                                                                              l 1
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plant will include cold criticality testing of the plant at es-sentially ambient temperature and atmospheric pressure.                           The power 1 vel during this phase of testing will be in the range of .0001% to .001% of rated power.         Details of the testing to l             be performed during this phase are described in the Notaro Af-fidavit.
plant will include cold criticality testing of the plant at es-sentially ambient temperature and atmospheric pressure.
1 (15)   The review of Chapter 15 revealed that of the.38 accident or transient events included there, 15 of the events could not occur because of the operating conditions of the plant during Phase I.         See Exhibit 2. A number of these events are not possible because the reactor will be at essentially ambient temperature and pressure and no steam will be gener-ated. For example, the generator load rejection event (FSAR
The power 1 vel during this phase of testing will be in the range of.0001% to.001% of rated power.
              $ 15A.1.1) could not occur during this testing phase because steam is needed to drive the main turbine generator to permit connecting it to the LILCO transmission system.                 Another exam-ple, the loss of condenser vacuum event (FSAR 5 15A.1.21),
Details of the testing to l
could not occur because it assumes that. steam is available to draw a vacuum in the main condenser.         A third example, the in-advertent HPCI pump start event (FSAR 5 15A.1.10), could not occur because there will be no steam available to power the HPCI pump, a steam driven ECCS pump.         Other Chapter 15 events could not occur because.they are precluded by the configuration
be performed during this phase are described in the Notaro Af-fidavit.
1 (15)
The review of Chapter 15 revealed that of the.38 accident or transient events included there, 15 of the events could not occur because of the operating conditions of the plant during Phase I.
See Exhibit 2.
A number of these events are not possible because the reactor will be at essentially ambient temperature and pressure and no steam will be gener-ated.
For example, the generator load rejection event (FSAR
$ 15A.1.1) could not occur during this testing phase because steam is needed to drive the main turbine generator to permit connecting it to the LILCO transmission system.
Another exam-ple, the loss of condenser vacuum event (FSAR 5 15A.1.21),
could not occur because it assumes that. steam is available to draw a vacuum in the main condenser.
A third example, the in-advertent HPCI pump start event (FSAR 5 15A.1.10), could not occur because there will be no steam available to power the HPCI pump, a steam driven ECCS pump.
Other Chapter 15 events could not occur because.they are precluded by the configuration e
m


_11 l
_11 of the plant during this phase of low po.wer testing.
  . of the plant during this phase of low po.wer testing.               An exam-ple of this type of event is the MSIV closure (FSAR 5 15A.1.4).
An exam-ple of this type of event is the MSIV closure (FSAR 5 15A.1.4).
The MSI s will normally be closed throughout all of the op-erations conducted during this phase of low power testing.                             In any event, there is no steam generated by the reactor to flow through the steam lines.
The MSI s will normally be closed throughout all of the op-erations conducted during this phase of low power testing.
(16)     In addition to the 15 events that could not occur during Phase I, many of the 23 events remaining in the Chapter 15 analysis are far less likely to occur during low power testing than during normal operations.       For example, the recirculation pump trip (FSAR $ 15A.1.20), the recirculation
In any event, there is no steam generated by the reactor to flow through the steam lines.
:      pump seizure (FSAR 5 15A.1.22), the recirculation flow control
(16)
In addition to the 15 events that could not occur during Phase I, many of the 23 events remaining in the Chapter 15 analysis are far less likely to occur during low power testing than during normal operations.
For example, the recirculation pump trip (FSAR $ 15A.1.20), the recirculation pump seizure (FSAR 5 15A.1.22), the recirculation flow control failures (FSAR 9 15A.1.23 and 24) and the abnormal startup of
)
)
failures (FSAR 9 15A.1.23 and 24) and the abnormal startup of idle recirculation pump (FSAR $ 15A.1.25) events, although physically possible, are not as likely to occur because the recirculation pumps are used for only limited periods of time during this phase of the testing program.         Similarly, the loss of feedwater event (FSAR $ 15A.1.18) is very unlikely because little, if any, make-up water will have to be supplied to the reactor. Moreover, make-up water would not normally be supplied by the feedwater system.under these conditions.                       Other very unlikely events include miscellaneous.small releases out-side primary containment (FSAR $ 15.1.29), off design l
idle recirculation pump (FSAR $ 15A.1.25) events, although physically possible, are not as likely to occur because the recirculation pumps are used for only limited periods of time during this phase of the testing program.
Similarly, the loss of feedwater event (FSAR $ 15A.1.18) is very unlikely because little, if any, make-up water will have to be supplied to the reactor.
Moreover, make-up water would not normally be supplied by the feedwater system.under these conditions.
Other very unlikely events include miscellaneous.small releases out-side primary containment (FSAR $ 15.1.29), off design l
l
l
                                                                                                        'l
'l
                                                                                                = . - .
=.


l operational transient as a consequence of instrument line fail-I ure (FSAR $ 15.1.30), and feedwater system piping break (FSAR
. l operational transient as a consequence of instrument line fail-I ure (FSAR $ 15.1.30), and feedwater system piping break (FSAR
            $ 15.1.3 ). Thus, many of the Chapter 15 events that are phys-ically possible during Phase II remain very unlikely in light of the plant conditions that will then exist.
$ 15.1.3 ).
(17) Nonetheless, all 23 possible events contained in Chapter 15 were reviewed to reaffirm that the consequences of these events, should one occur during Phase I of low power testing, would be bounded by the consequences analyzed for the event considered in the FSAR. A discussion of some of the 23 possible events contained in Chapter 15 illustrates the basis for this conclusion. The continuous control rod withdrawal during startup event (FSAR $ 15A.1.12) is applicable to op-eration in the power, source and/or intermediate range of op-eration. During cold functional criticality testing, the reac-tor will operate in the source and intermediate ranges and therefore the conclusions contained in Chapter 15 are applica-ble to this event should it occur during this phase of low power testing. As the FSAR indicates, this event would not re-sult in any rele.e ' af radioactive material from the fuel at any power level. Another example is the fuel handling accident (FSAR $ 15.1.36). As stated in the FSAR, the most severe fuel handling accident from a radiological viewpoint is a dropping l
Thus, many of the Chapter 15 events that are phys-ically possible during Phase II remain very unlikely in light of the plant conditions that will then exist.
r -  --  r     v-- , , - - ,- , ,--<- 7 -__..,e - --            -
(17)
o
Nonetheless, all 23 possible events contained in Chapter 15 were reviewed to reaffirm that the consequences of these events, should one occur during Phase I of low power testing, would be bounded by the consequences analyzed for the event considered in the FSAR.
A discussion of some of the 23 possible events contained in Chapter 15 illustrates the basis for this conclusion.
The continuous control rod withdrawal during startup event (FSAR $ 15A.1.12) is applicable to op-eration in the power, source and/or intermediate range of op-eration.
During cold functional criticality testing, the reac-tor will operate in the source and intermediate ranges and therefore the conclusions contained in Chapter 15 are applica-ble to this event should it occur during this phase of low power testing.
As the FSAR indicates, this event would not re-sult in any rele.e ' af radioactive material from the fuel at any power level.
Another example is the fuel handling accident (FSAR $ 15.1.36).
As stated in the FSAR, the most severe fuel handling accident from a radiological viewpoint is a dropping o
r r
v--
7
-__..,e


_ = -         .                      . . ._  . - . -        - -        .
_ = -
of the fuel assembly onto the top of the core.                           The FSAR analy-l sis assumes that the fuel contains a fission product inventory                                         i l       equivalent to operation of 1000 days at full rated power.                                       This .
- of the fuel assembly onto the top of the core.
;        assumption results in an equilibrium fission product concentra-I tion at the time the reactor is shut down.                       But as already noted, the fission product inventories in the core will be sig-nificantly less during Phase II low power testing than the in-ventories analyzed in the FSAR because of the extremely low power levels (.0001% to .001% of rated power) achieved during this testing.                 Thus, even if a handling accident took place and fuel damage did occur, there would be significantly less fis-
The FSAR analy-l sis assumes that the fuel contains a fission product inventory i
;        sion products to be released from the core than those that have
l equivalent to operation of 1000 days at full rated power.
}         already been analyzed and found acceptable in the FSAR.                                   A third example is the liquid radwaste tank rupture event (FSAR $
This assumption results in an equilibrium fission product concentra-I tion at the time the reactor is shut down.
15.1.32).                 This event assumes the rupture of a radwaste tank that contains a substantial amount of contaminated liquids gen-erated during the operation of the reactor.                       But again, since Phase II low power testing results in insignificant power lev-els in the reactor, there will be little, if any, radioactive i         liquids in the radwaste tank should such a rupture occur.
But as already noted, the fission product inventories in the core will be sig-nificantly less during Phase II low power testing than the in-ventories analyzed in the FSAR because of the extremely low power levels (.0001% to.001% of rated power) achieved during this testing.
Thus, even if a handling accident took place and fuel damage did occur, there would be significantly less fis-sion products to be released from the core than those that have
}
already been analyzed and found acceptable in the FSAR.
A third example is the liquid radwaste tank rupture event (FSAR $
15.1.32).
This event assumes the rupture of a radwaste tank that contains a substantial amount of contaminated liquids gen-erated during the operation of the reactor.
But again, since Phase II low power testing results in insignificant power lev-els in the reactor, there will be little, if any, radioactive i
liquids in the radwaste tank should such a rupture occur.
I
I
]         Thus, even the minimal consequences already described in the
]
        'FSAR for the design basis event would be further reduced under these low power testing conditions.                     For each of these events, the review concluded that the consequences are significantly 1
Thus, even the minimal consequences already described in the
                - . - ,    . - . ,                      ,.e   -    -        .%.                T T
'FSAR for the design basis event would be further reduced under these low power testing conditions.
For each of these events, the review concluded that the consequences are significantly 1
,.e T
T


less severe for any event occurring during the cold functional criticality testing than for the event analyzed in Chapter 15.
, less severe for any event occurring during the cold functional criticality testing than for the event analyzed in Chapter 15.
To summa ize, because of the extremely low power levels reached during this' testing phase, fission product inventory in the core will be only a small fraction of that assumed for the Chapter 15 analyses. As indicated above, the FSAR assumes op-eration at 100% power for 1000 days in calculating fission product inventory; the inventory during Phase II low power testing will be less than one one-hundred-thousandth (.00001) of the fission product inventory assumed in the FSAR. Conse-quently, none of the events analyzed in Chapter 15 could result in a release of radioactivity during cold criticality testing that would harm the public health and safety.
To summa ize, because of the extremely low power levels reached during this' testing phase, fission product inventory in the core will be only a small fraction of that assumed for the Chapter 15 analyses.
(18) The review of Chapter 15 events for Phase II testing and the conclusions reached are unaffected by any unavailability of the TDI diesels. Of the 23 possible Chapter 15 events reviewed, 20 of the events in the FSAR do not require the assumption of loss or unavailability of offsite AC power.
As indicated above, the FSAR assumes op-eration at 100% power for 1000 days in calculating fission product inventory; the inventory during Phase II low power testing will be less than one one-hundred-thousandth (.00001) of the fission product inventory assumed in the FSAR.
;      See Exhibit 2. Thus, our conclusions for these 20 of the 23 possible events are independent of the status of the diesels.
Conse-quently, none of the events analyzed in Chapter 15 could result in a release of radioactivity during cold criticality testing that would harm the public health and safety.
(19) The three events that do assume loss or unavailability of offsite AC power are (1) pipe breaks inside the primary containment (LOCA) (FSAR $ 15.1.34), (2) feedwater
(18)
The review of Chapter 15 events for Phase II testing and the conclusions reached are unaffected by any unavailability of the TDI diesels.
Of the 23 possible Chapter 15 events reviewed, 20 of the events in the FSAR do not require the assumption of loss or unavailability of offsite AC power.
See Exhibit 2.
Thus, our conclusions for these 20 of the 23 possible events are independent of the status of the diesels.
(19)
The three events that do assume loss or unavailability of offsite AC power are (1) pipe breaks inside the primary containment (LOCA) (FSAR $ 15.1.34), (2) feedwater


__ _ _ . ~       . . . _  _ _ _ . _-_ _. ___          _      __  -      . _ _ _ ._
__ _ _. ~
4 system piping break (FSAR $ 15.1.37), and (3) the loss of AC power event (FSAR $ 15A.1.19).               With respect to these~ events, I                 the LOCI'would be the most limiting event.                 The review has           -
4 system piping break (FSAR $ 15.1.37), and (3) the loss of AC power event (FSAR $ 15A.1.19).
shown that'if a LOCA did occur during the cold criticality testing phase, however remote that possibility, there would be time on the order of months available to restore make-up water for core cooling.           At the power levels achieved during Phase l                 II, fission product inventory is very low.                 At most, decay heat will, on the average, be a fraction of a watt per rod, with no single rod exceeding approximately 2 watts.                 This is less, roughly, than the heat output of a Christmas tree bulb.                       It 4
With respect to these~ events, I
the LOCI'would be the most limiting event.
The review has shown that'if a LOCA did occur during the cold criticality testing phase, however remote that possibility, there would be time on the order of months available to restore make-up water for core cooling.
At the power levels achieved during Phase l
II, fission product inventory is very low.
At most, decay heat will, on the average, be a fraction of a watt per rod, with no single rod exceeding approximately 2 watts.
This is less, roughly, than the heat output of a Christmas tree bulb.
It 4
follows that the fuel cladding temperature would not exceed the limits of 10 CFR $ 50.46 and Appendix K even after months with-out cooling and without any source of AC power.
follows that the fuel cladding temperature would not exceed the limits of 10 CFR $ 50.46 and Appendix K even after months with-out cooling and without any source of AC power.
(20)     The loss of AC power event (FSAR $ 15A.1.19) and the feedwater system piping break (FSAR 5 15.1.37) under cold criticality testing conditions do not rely on the diesel gener-ators for mitigation of the event.               For these events, since no loss of coolant occurs and the decay heat is minimal, core cooling is achieved. without AC power, using the existing core water inventory and heat losses to ambient, for essentially un-limited periods of time.               In any event, as demonstrated in the Schiffmacher Affidavit, AC power sources can and will be
(20)
The loss of AC power event (FSAR $ 15A.1.19) and the feedwater system piping break (FSAR 5 15.1.37) under cold criticality testing conditions do not rely on the diesel gener-ators for mitigation of the event.
For these events, since no loss of coolant occurs and the decay heat is minimal, core cooling is achieved. without AC power, using the existing core water inventory and heat losses to ambient, for essentially un-limited periods of time.
In any event, as demonstrated in the Schiffmacher Affidavit, AC power sources can and will be


readily supplied to the Shoreham plant even if one assumes the simultaneous loss of all three emergency diesel generators.
. readily supplied to the Shoreham plant even if one assumes the simultaneous loss of all three emergency diesel generators.
(21)   In addition to our conclusions that the limiting LOCA event could not approach the limits of 10 CFR $ 50.46 and Appendix K during Phase II low power testing, there are other reasons why our findings with respect to the three events that assume loss of AC power are independent of the availability of the TDI diesels. The LOCA (pipe break inside containment) and the feedwater system piping break postulate the double ended rupture of a piping system. Because the reactor will be at es-sentially ambient temperature and atmospheric pressure during Phase II, it is extremely unlikely that such a pipe break would ever occur. In fact, the NRC Staff does not require double ended ruptures to be postulated for low temperature and low pressure systems in safety analyses. Thus, these events are much less likely during cold criticality testing than during normal operation.
(21)
(22) The review of Chapter 15 events for cold criti-cality testing indicates that performance of these activities at Shoreham involves essentially no risk to the public health and safety. This conclusion is not affected by any postulated   ,
In addition to our conclusions that the limiting LOCA event could not approach the limits of 10 CFR $ 50.46 and Appendix K during Phase II low power testing, there are other reasons why our findings with respect to the three events that assume loss of AC power are independent of the availability of the TDI diesels.
1 diesel unavailability. In fact, even if AC power were not
The LOCA (pipe break inside containment) and the feedwater system piping break postulate the double ended rupture of a piping system.
-                                                                            )
Because the reactor will be at es-sentially ambient temperature and atmospheric pressure during Phase II, it is extremely unlikely that such a pipe break would ever occur.
i available for extended periods of time, fuel design limits and
In fact, the NRC Staff does not require double ended ruptures to be postulated for low temperature and low pressure systems in safety analyses.
Thus, these events are much less likely during cold criticality testing than during normal operation.
(22)
The review of Chapter 15 events for cold criti-cality testing indicates that performance of these activities at Shoreham involves essentially no risk to the public health and safety.
This conclusion is not affected by any postulated diesel unavailability.
In fact, even if AC power were not i
available for extended periods of time, fuel design limits and


design conditions of the reactor coolant pressure boundary would not be approached or exceeded as a result of anticipated operati nal occurrences, and the core would be adequately cooled in the unlikely event of a postulated accident.
. design conditions of the reactor coolant pressure boundary would not be approached or exceeded as a result of anticipated operati nal occurrences, and the core would be adequately cooled in the unlikely event of a postulated accident.
Phases III and IV:
Phases III and IV:
Low Power Testing Up To 5% of Rated Power (23)   These aspects of low power testing will include operation of the plant at power levels up to 5% of rated power.
Low Power Testing Up To 5% of Rated Power (23)
These aspects of low power testing will include operation of the plant at power levels up to 5% of rated power.
Details of the testing to be performed during this phase of op-eration are described as Phases III and IV in the Notaro Affi-davit.
Details of the testing to be performed during this phase of op-eration are described as Phases III and IV in the Notaro Affi-davit.
(24) The review of the 38 Chapter 15 events for these phases of low power testing operations revealed that two of the events in Chapter 15, generator load rejection (FSAR 5 15A.1.1) and turbine trip with generator breaker failure (FSAR 5 15.1.2) cannot occur because the generator will not be connected to the grid during these phases of testing. A third event, the cask drop, is precluded by design as stated in FSAR 5 15.1.28. See Exhibit 3.
(24)
(25) Of the remaining 35 events that can occur during this phase of operation, 31 of the events do not assume loss or unavailability of AC power. For each of these 31 events,
The review of the 38 Chapter 15 events for these phases of low power testing operations revealed that two of the events in Chapter 15, generator load rejection (FSAR 5 15A.1.1) and turbine trip with generator breaker failure (FSAR 5 15.1.2) cannot occur because the generator will not be connected to the grid during these phases of testing.
A third event, the cask drop, is precluded by design as stated in FSAR 5 15.1.28.
See Exhibit 3.
(25)
Of the remaining 35 events that can occur during this phase of operation, 31 of the events do not assume loss or unavailability of AC power.
For each of these 31 events,


operation of the plant up to 5% of rated power will be bounded by the Chapter 15 analysis.                             Since the Chapter 15 analysis con-i'              e, siders all possible phases of plant operation, it follows that operation a't 5% can result in consequences ~1ess severe than those analyzed in Chapter 15.                             For example, the turbine trip event (FSAR $ 15A.1.2) assumes that the limiting event occurs with the reactor operating at 105% of rated steam flow coupled with failure of the turbine bypass valves to open.                                           Even this limiting event does not result in any fuel failures.                                             ESAR 5 15A.1.2 specifically notes that turbine trips at power levels less than 30% of rated power are bounded by the limiting analy-sis. Another example is the loss of feedwater heating event (FSAR $ 15A.1.8).                         This event assumes continuous operation of the feedwater system and the most severe possible loss of feedwater heating, resulting in the injection of colder feedwater.       For operation at power levels less than 5%, the im-pact of lost feedwater heating is minimal because of the low feedwater flow.                         Since these analyses are not required to as-sume the absence of AC power, potential unavailability of the
. operation of the plant up to 5% of rated power will be bounded by the Chapter 15 analysis.
,        TDI diesels has no effect on the assessment of these events.
Since the Chapter 15 analysis con-i e,
l (26)           Not only are the results of these 31 events bounded by the Chapter 15 analysis, the consequences of these events are also less than the consequences stated in the FSAR.
siders all possible phases of plant operation, it follows that operation a't 5% can result in consequences ~1ess severe than those analyzed in Chapter 15.
l l
For example, the turbine trip event (FSAR $ 15A.1.2) assumes that the limiting event occurs with the reactor operating at 105% of rated steam flow coupled with failure of the turbine bypass valves to open.
Even this limiting event does not result in any fuel failures.
ESAR 5 15A.1.2 specifically notes that turbine trips at power levels less than 30% of rated power are bounded by the limiting analy-sis.
Another example is the loss of feedwater heating event (FSAR $ 15A.1.8).
This event assumes continuous operation of the feedwater system and the most severe possible loss of feedwater heating, resulting in the injection of colder feedwater.
For operation at power levels less than 5%, the im-pact of lost feedwater heating is minimal because of the low feedwater flow.
Since these analyses are not required to as-sume the absence of AC power, potential unavailability of the TDI diesels has no effect on the assessment of these events.
l (26)
Not only are the results of these 31 events bounded by the Chapter 15 analysis, the consequences of these events are also less than the consequences stated in the FSAR.
l


1
1
Line 164: Line 316:
i 1
i 1
First, the power limitations during low power testing up to 5%
First, the power limitations during low power testing up to 5%
power, the fission product inventory in the core will not ex-4 ceed 5%b of the values assumed in the FSAR.       In fact, because of   -
power, the fission product inventory in the core will not ex-4 b
;        the intermittant type of operations conducted during low power f
ceed 5% of the values assumed in the FSAR.
In fact, because of the intermittant type of operations conducted during low power f
testing, equilibrium fission product inventory for even 5%
testing, equilibrium fission product inventory for even 5%
i power is unlikely to be achieved.         This low fission product in-ventory reduces risk in two ways:         (a) the amount of decay heat present in the core following shutdown is substantially re-duced, and (b) the amount of radioactivity that could be re-leased upon fuel failure is substantially reduced.
i power is unlikely to be achieved.
(27) The second factor contributing to the signifi-
This low fission product in-ventory reduces risk in two ways:
.        cantly lower risk during low power operation is the increased I
(a) the amount of decay heat present in the core following shutdown is substantially re-duced, and (b) the amount of radioactivity that could be re-leased upon fuel failure is substantially reduced.
time available for preventive or mitigating action should such l       action be deemed desirable by the operator.         Longer time is available because the limited power levels mean that it takes longer for the plant to reach setpoints and limits.         For exam-ple, on loss of feedwater (FSAR 5 15A.1.18), the water level in the reactor will decrease at a slower rate than if the event occured at 100% power. This gives the operator more time to act manually to restore feedwater before an automatic action takes place. Similarly, in the loss of condenser vacuum event (FSAR $ 15.A.1.21), the operator will have more time to identi-fy the decreasing vacuum and to take steps to remedy the i
(27)
The second factor contributing to the signifi-cantly lower risk during low power operation is the increased I
time available for preventive or mitigating action should such l
action be deemed desirable by the operator.
Longer time is available because the limited power levels mean that it takes longer for the plant to reach setpoints and limits.
For exam-ple, on loss of feedwater (FSAR 5 15A.1.18), the water level in the reactor will decrease at a slower rate than if the event occured at 100% power.
This gives the operator more time to act manually to restore feedwater before an automatic action takes place.
Similarly, in the loss of condenser vacuum event (FSAR $ 15.A.1.21), the operator will have more time to identi-fy the decreasing vacuum and to take steps to remedy the i
i l
i l


situation before automatic actions such as turbine trip, feedpump trip or main steam isolation occur.       Another example is the   in steam isolation valve closure event (FSAR 5 15A.1.4).'   At five percent power, the amount of heat produced upon isolation of the reactor vessel (which is followed by a reactor scram) results in a much slower pressure and tempera-ture increase than would be experienced at 100% power.       This gives the operator more time to manually initiate reactor cool-ing rather than relying on automatic action.       In effect, the operator may end the transient before there is any substantial impact on the plant.
. situation before automatic actions such as turbine trip, feedpump trip or main steam isolation occur.
(28)       The third factor contributing to the signifi-cantly lower risk during low power testing is the reduction in the required capacity for mitigating systems.       Because of the lower levels of decay heat present following operation at 5%
Another example is the in steam isolation valve closure event (FSAR 5 15A.1.4).'
power, the demand for core cooling and auxiliary systems is substantially reduced, permitting the operation of fewer sys-tems a'nd components to mitigate any event.       It follows that the AC power requirements for event mitigation are substantially reduced for 5% power operation as compared to 100% power op-eration.
At five percent power, the amount of heat produced upon isolation of the reactor vessel (which is followed by a reactor scram) results in a much slower pressure and tempera-ture increase than would be experienced at 100% power.
(29)       As already noted, only four of the events ana-lyzed in Chapter 15 require the assumption of the                     I
This gives the operator more time to manually initiate reactor cool-ing rather than relying on automatic action.
In effect, the operator may end the transient before there is any substantial impact on the plant.
(28)
The third factor contributing to the signifi-cantly lower risk during low power testing is the reduction in the required capacity for mitigating systems.
Because of the lower levels of decay heat present following operation at 5%
power, the demand for core cooling and auxiliary systems is substantially reduced, permitting the operation of fewer sys-tems a'nd components to mitigate any event.
It follows that the AC power requirements for event mitigation are substantially reduced for 5% power operation as compared to 100% power op-eration.
(29)
As already noted, only four of the events ana-lyzed in Chapter 15 require the assumption of the


l' unavailability of offsite AC power for operation during Phases I
. unavailability of offsite AC power for operation during Phases I
j                   III and IV.       Of these four events, the loss of coolant accident
j III and IV.
;                              e is the $ost limiting event.       The Chapter 15 LOCA analysis as-sumes the unavailability of offsite AC power.         This is a con-servative licensing assumption.         In fact, as described in de-tail in the Schiffmacher Affidavit, there are multiple sources I
Of these four events, the loss of coolant accident e
of AC power available to the Shoreham site (e.g., emergency diesel generators, two normal sources of offsite power,
is the $ost limiting event.
;                  blackstart gas turbines at Holtsville, Southhold, and East Hampton, a blackstart gas turbine on the Shoreham site, and mo-bile diesel generators).       Thus, AC power will be available at j                   Shoreham to mitigate a loss of coolant accident during low power operations up to 5% rated power.         In the unlikely event offsite AC power is lost, it can be restored within sufficient time to prevent exceeding the limits of 10 CFR 5 50.46 and Ap-pendix K.       GE has determined that for 5% power so long as reflooding of the core has occurred within approximately one hour, 5 50.46 criteria will be met.2/         As the Schiffmacher Af-fidavit demonstrates, power can be restored to Shoreham within minutes.       An evaluation has been performed to assure the ade-quacy of containment isolation in the event AC power sources 2/   As shown in the Exhibit 4 below, lower power levels will result in more time to restore power and core cooling for a postulated LOCA. Thus, for 1% power approximately 5 hours are available.
The Chapter 15 LOCA analysis as-sumes the unavailability of offsite AC power.
This is a con-servative licensing assumption.
In fact, as described in de-tail in the Schiffmacher Affidavit, there are multiple sources I
of AC power available to the Shoreham site (e.g.,
emergency diesel generators, two normal sources of offsite power, blackstart gas turbines at Holtsville, Southhold, and East Hampton, a blackstart gas turbine on the Shoreham site, and mo-bile diesel generators).
Thus, AC power will be available at j
Shoreham to mitigate a loss of coolant accident during low power operations up to 5% rated power.
In the unlikely event offsite AC power is lost, it can be restored within sufficient time to prevent exceeding the limits of 10 CFR 5 50.46 and Ap-pendix K.
GE has determined that for 5% power so long as reflooding of the core has occurred within approximately one hour, 5 50.46 criteria will be met.2/
As the Schiffmacher Af-fidavit demonstrates, power can be restored to Shoreham within minutes.
An evaluation has been performed to assure the ade-quacy of containment isolation in the event AC power sources 2/
As shown in the Exhibit 4 below, lower power levels will result in more time to restore power and core cooling for a postulated LOCA.
Thus, for 1% power approximately 5 hours are available.
i
i


    - - .        ._- -_ - . . _ - = _ _ _ _ . - -     _
-_ -.. _ - = _ _ _ _. - -
l
l
\
\\
l                                                                                                                                                     d cannot provide immediate isolation in a LOCA.                                                 Based upon the results of this evaluation, we have concluded that through the                                                                           !
- l d
!            use of appropriate manual action, containment isolation can be
cannot provide immediate isolation in a LOCA.
;            accomplished in a timely manner.
Based upon the results of this evaluation, we have concluded that through the use of appropriate manual action, containment isolation can be accomplished in a timely manner.
I (30)                 For the other three events, (1) loss of AC power l
I (30)
l           (FSAR $ 15A.1.19), (2) pipe break outside containment (PBOC)
For the other three events, (1) loss of AC power l
J (steam line break accident) (FSAR 6 15.1.35) and (3) feedwater system piping break (FSAR $ 15.1.37), the reactor would auto-matically isolate.                       This isolation is not dependent upon the availability of AC power.                             For all three events, both HPCI and RCIC would be available to provide reactor coolant makeup.
l (FSAR $ 15A.1.19), (2) pipe break outside containment (PBOC)
Given the heat capacity of passive heat sinks such as structur-4 al steel, suppression pool cooling would not be required for about 30 days.                       Therefore, there is ample time for AC power to i             be restored.                     Furthermore, assuming loss of offsite power in the context of pipe breaks outside containment (main steam line break accident and feedwater system break accident) is a con-servatism which stems from the PBOC analysis methodology.                                                                 That methodology requires the assumption of a loss of offsite power for pipe breaks which result directly in a plant trip of the turbine generator system or reactor protection system.                                                               Not-withstanding grid stability analyses, it is assumed that plant trips could cause perturbations of the grid, resulting in the
J (steam line break accident) (FSAR 6 15.1.35) and (3) feedwater system piping break (FSAR $ 15.1.37), the reactor would auto-matically isolate.
                                                          - y --- -                  %y v q v         ,y --.w ,.,-.-y .
This isolation is not dependent upon the availability of AC power.
9-$.,e     r*
For all three events, both HPCI and RCIC would be available to provide reactor coolant makeup.
Given the heat capacity of passive heat sinks such as structur-4 al steel, suppression pool cooling would not be required for about 30 days.
Therefore, there is ample time for AC power to i
be restored.
Furthermore, assuming loss of offsite power in the context of pipe breaks outside containment (main steam line break accident and feedwater system break accident) is a con-servatism which stems from the PBOC analysis methodology.
That methodology requires the assumption of a loss of offsite power for pipe breaks which result directly in a plant trip of the turbine generator system or reactor protection system.
Not-withstanding grid stability analyses, it is assumed that plant trips could cause perturbations of the grid, resulting in the e
w-e y
y 3
%y v q v
,y
--.w 9
,.,-.-y 9-$.,e r*


loss of offsite power. For operation at 5% power or less, how-ever, the turbine generator is not connected to the grid, and therefo e any assumption of induced perturbation to the offsite grid is not valid.
loss of offsite power.
(31) Based on our review of Chapter 15, operation of the plant during low power testing up to levels of 5% of rated power poses no undue risk to the public health and safety. In fact, any risk is substantially less than that already found to be acceptable by the NRC Staff in its review of Chapter 15.
For operation at 5% power or less, how-ever, the turbine generator is not connected to the grid, and therefo e any assumption of induced perturbation to the offsite grid is not valid.
(31)
Based on our review of Chapter 15, operation of the plant during low power testing up to levels of 5% of rated power poses no undue risk to the public health and safety.
In fact, any risk is substantially less than that already found to be acceptable by the NRC Staff in its review of Chapter 15.
Even if the Shoreham TDI diesels are assumed to be unavailable, there is ample assurance that fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded as a result of anticipated operational occurrences, and that the core will be cooled and containment integrity and other vital functions will be maintained in the event of any postulated accident.
Even if the Shoreham TDI diesels are assumed to be unavailable, there is ample assurance that fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded as a result of anticipated operational occurrences, and that the core will be cooled and containment integrity and other vital functions will be maintained in the event of any postulated accident.
Glenn G. Sherwood M         Gd Atambir S. Rao Eugene C. Eckert
Glenn G. Sherwood M
Gd Atambir S.
Rao Eugene C.
Eckert


                ._-      - - .        _ .. - , _ -        .  . _ __ .. . - .                      - . _ _ _ . . _ - = . _ _ .
. _ - =. _ _.
i i
i i i
i STATE OF   MEtuNyc)
MEtuNyc)
                                  )  To-wit:
STATE OF
:          COUNTY GE lYES7 tee 5722)                                                                                       ,
)
To-wit:
COUNTY GE lYES7 tee 5722)
Subscribed to before me this h ay of March, 1984.
Subscribed to before me this h ay of March, 1984.
n"u eM Notary p lic JOSE M. TEJAD1 Notary Pubhc. State et r;e., y.n My commission expires:                                 Ns. 03 4727173 yualihed in Srcas County Cert. Filed in Westchester (a -
n"u eM Notary p lic JOSE M. TEJAD1 Notary Pubhc. State et r;e., y.n My commission expires:
Ns. 03 4727173 yualihed in Srcas County Cert. Filed in Westchester (a -
Commission Empires March 30,19[fy.
Commission Empires March 30,19[fy.
i 1
i 1
.1 i
.1 i
a                     -        -                            ,            .    -                        n
a n


                            ~
~
Exhibit 1
Exhibit 1
                    , FUEL LOAD AND PRECRITICALITY TESTING Chapter 15 Event                   Event Category
, FUEL LOAD AND PRECRITICALITY TESTING Chapter 15 Event Event Category 1.
: 1. Generator Load Rejection                         *
Generator Load Rejection 2.
-        2. Turbine Trip                                     *
Turbine Trip 3.
: 3. Turbine Trip with Failure of Generator Breakers to Open                       *
Turbine Trip with Failure of Generator Breakers to Open 4.
: 4. MSIV Closure                                     *
MSIV Closure 5.
: 5. Pressure Regulator Failure - Open               *
Pressure Regulator Failure - Open 6.
: 6. Pressure Regulator Failure - Closed             *
Pressure Regulator Failure - Closed 7.
: 7. Feedwater Controller Failure -
Feedwater Controller Failure -
Maximum Demand                                   ***
Maximum Demand 8.
: 8. Loss of Feedwater Heating                       *
Loss of Feedwater Heating 9.
: 9. Shutdown Cooling (RHR) Malfunction -
Shutdown Cooling (RHR) Malfunction -
Decreasing Temperature                           ***
Decreasing Temperature 10.
: 10. Inadvertent HPCI Pump Start                       *
Inadvertent HPCI Pump Start 11.
: 11. Continuous Control Rod Withdrawal During Power Range Operation                     *
Continuous Control Rod Withdrawal During Power Range Operation Event not possible.
* Event not possible.
Component operation possible but Chapter 15 phenomena cannot occur.
        ** Component operation possible but Chapter 15 phenomena cannot occur.
Event possible but no consequences.
        *** Event possible but no consequences.
-l l
                                                                              -l 1
l l


~
~
: 12. Continuous Rod Withdrawal During Reactor Startup                       ***
. 12.
: 13. Control Rod Removal Error During Rdfueling                             ***
Continuous Rod Withdrawal During Reactor Startup 13.
: 14. Fuel-Assembly Insertion Error During Refueling                       ***
Control Rod Removal Error During Rdfueling 14.
: 15. Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location                               *
Fuel-Assembly Insertion Error During Refueling 15.
: 16. Inadvertent Loading and Operation of a Fuel Assembly in Improper Location                               *
Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location 16.
: 17. Inadvertent Opening of a Safety / Relief Valve                 *
Inadvertent Loading and Operation of a Fuel Assembly in Improper Location 17.
: 18. Loss of Feedwater Flow                 ***
Inadvertent Opening of a Safety / Relief Valve 18.
: 19. Loss of AC Power                       ***
Loss of Feedwater Flow 19.
: 20. Recirculation Pump Trip               **
Loss of AC Power 20.
: 21. Loss of Condenser Vacuum               *
Recirculation Pump Trip 21.
: 22. Recirculation Pump Seizure             **
Loss of Condenser Vacuum 22.
: 23. Recirculation Flow Control Failure -
Recirculation Pump Seizure 23.
Decreasing Flow                       **
Recirculation Flow Control Failure -
: 24. Recirculation Flow Control Failure With Increasing Flow                   **
Decreasing Flow 24.
: 25. Abnormal Startup of Idle Recirculation Pump                     **
Recirculation Flow Control Failure With Increasing Flow 25.
: 26. Core Coolant Temperature Increase     ***
Abnormal Startup of Idle Recirculation Pump 26.
: 27. Anticipated Transients Without SCRAM (ATWS)                           *
Core Coolant Temperature Increase 27.
: 28. Cask Drop Accident                     *      ;
Anticipated Transients Without SCRAM (ATWS) 28.
1
Cask Drop Accident 29.
: 29. Miscellaneous Small Releases                   l Outside Primary Containment           ***
Miscellaneous Small Releases Outside Primary Containment l
l


3-
3-30.
: 30. Off Design Operational Transient as a Consequence of Instrument Line Failure                                                                             ***
Off Design Operational Transient as a Consequence of Instrument Line Failure 31.
: 31. Main Condenser Gas Treatment                                                                                                     -
Main Condenser Gas Treatment Syste,m Failure 32.
Syste,m Failure                                                                         *
Liquid Radwaste Tank Rupture 33.
: 32. Liquid Radwaste Tank Rupture                                                             ***
Control Rod Drop Accident 34.
: 33. Control Rod Drop Accident                                                               ***
Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident) 35.
: 34. Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)                                                   ***
Pipe Breaks Outside Primary Containment (Steam Line Break Accident) 36.
: 35. Pipe Breaks Outside Primary Containment (Steam Line Break Accident)                                                 *
Fuel Handling Accident 37.
: 36. Fuel Handling Accident                                                                   ***
Feedwater System Piping Break 38.
: 37. Feedwater System Piping Break                                                           ***
Failure of Air Ejector Lines
: 38. Failure of Air Ejector Lines
_____.____-_._._._______m
* _____.____-_._._._______m     _-_____-___-_____-_______-_____m.._   -..-___-_ _--__
_-_____-___-_____-_______-_____m.._
_              _____m____._-__m______   _ _ ______-_ ______.
_____m____._-__m______


Exhibit 2 COLD CRITICALITY TESTING Assumes Un-Event     availability Chapter 15 Event                   Category   of Offsite AC
Exhibit 2 COLD CRITICALITY TESTING Assumes Un-Event availability Chapter 15 Event Category of Offsite AC N/A 1.
: 1. Generator Load Rejection                                 N/A
Generator Load Rejection N/A 2.
: 2. Turbine Trip                                             N/A
Turbine Trip 3.
: 3. Turbine Trip with Failure of Generator Breakers to Open
Turbine Trip with Failure of Generator Breakers to Open N/A 4.
* N/A
MSIV Closures N/A 5.
: 4. MSIV Closures                                           N/A 1      5. Pressure Regulator Failure - Open                       N/A Pressure Regulator Failure - Closed         *
Pressure Regulator Failure - Open N/A 1
: 6.                                                          N/A
6.
                          ~
Pressure Regulator Failure - Closed N/A
: 7. Feedwater Controller Failure -
~
Maximum Demand                               **          No Loss of Feedwater Heating                   *
7.
: 8.                                                          N/A
Feedwater Controller Failure -
: 9. Shutdown Cooling (RHR) Malfunction -
Maximum Demand No 8.
Decreasing Temperature                       **          No Inadvertent HPCI Pump Start                 *
Loss of Feedwater Heating N/A 9.
: 10.                                                          N/A
Shutdown Cooling (RHR) Malfunction -
: 11. Continuous Control Rod Withdrawal During Power Range Operation                             N/A
Decreasing Temperature No 10.
* Event not possible.                                                     ;
Inadvertent HPCI Pump Start N/A 11.
      **  Event possible but. essentially no consequences.
Continuous Control Rod Withdrawal During Power Range Operation N/A Event not possible.
                                - -  -e     ,            -,
Event possible but. essentially no consequences.
-e


l i       -
l i
                                                    ,             12. Continuous Rod Withdrawal During Reactor Startup                                                 **                                        No l
, 12.
l             13. Control Rod Removal Error During Refueling                                                       **                                      No
Continuous Rod Withdrawal During No Reactor Startup l
: 14. Fuel Assembly Insertion Error During Refueling                                               **                                        No
l 13.
            .15. Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location                                                         **                                      No
Control Rod Removal Error During No Refueling 14.
: 16. Inadvertent Loading and Operation of a Fuel Assembly in Improper Location                                                         **                                      No
Fuel Assembly Insertion Error No During Refueling
: 17. Inadvertent Opening of a Safety / Relief Valve                                                                                     N/A
.15.
: 18. Loss of Feedwater Flow                                           **                                        No
Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location No 16.
: 19. Loss of AC Power                                                 **                                        Yes a            20. Recirculation Pump Trip                                         **                                        No Loss of Condenser Vacuum                                         *
Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.
: 21.                                                                                                                N/A
Inadvertent Opening of a N/A Safety / Relief Valve 18.
: 22. Recirculation Pump Seizure                                       **                                        No
Loss of Feedwater Flow No 19.
: 23. Recirculation Flow Control Failure -
Loss of AC Power Yes 20.
Decreasing Flow                                                 **                                        No
Recirculation Pump Trip No a
: 24. Recirculation Flow Control Failure With Increasing Flow                                             **                                        No 1
21.
: 25. Abnormal Startup of Idle Recirculation Pump                                           -**                                          No
Loss of Condenser Vacuum N/A 22.
: 26. Core Coolant Temperature Increase                               **                                        No
Recirculation Pump Seizure No 23.
: 27. Anticipated Transients Without SCRAM (ATWS)                                                     **                                        No Cask Drop Accident                                               *
Recirculation Flow Control Failure -
: 28.                                                                                                                  N/A
Decreasing Flow No 24.
: 29. Miscellaneous Small Releases outside Primary Containment                                     **                                          No-l
Recirculation Flow Control Failure With Increasing Flow No 1
25.
Abnormal Startup of Idle Recirculation Pump No 26.
Core Coolant Temperature Increase No 27.
Anticipated Transients Without SCRAM (ATWS)
No 28.
Cask Drop Accident N/A 29.
Miscellaneous Small Releases outside Primary Containment No-


    .   .                                                                 l 1
. 30.
: 30. Off Design Operational Transient as a Consequence of Instrument                               l Line Failure                             **        No Mk'in Condenser Gas Treatment
Off Design Operational Transient as a Consequence of Instrument Line Failure No 31.
                  ~
Mk'in Condenser Gas Treatment
: 31.                                                          -
~
System Failure
System Failure N/A 32.
* N/A
Liquid Radwate Tank Rupture No 33.
: 32. Liquid Radwate Tank Rupture               **        No
Control Rod Drop Accident No 34.
: 33. Control Rod Drop Accident                 **        No
Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)
: 34. Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)   **        Yes
Yes 35.
: 35. Pipe Breaks Outside Primary Containment (steam line break accident)
Pipe Breaks Outside Primary Containment (steam line break accident)
* N/A
* N/A 36.
: 36. Fuel Handling Accident                   **        No
Fuel Handling Accident No 37.
: 37. Feedwater System Piping Break             **        Yes
Feedwater System Piping Break Yes 38.
: 38. Failure of Air Ejector Lines
Failure of Air Ejector Lines N/A i
* N/A i
f
f
                                              .,7 .._,. _-            .
.,7


              .                                                          I Exhibit 3
I Exhibit 3 5% POWER Assumes Un-Event availability Chapter 15 Event Category of Offsite AC N/A 1.
                  .              5% POWER Assumes Un-Event     availability Chapter 15 Event                 Category   of Offsite AC
Generator Load Rejection 2.
: 1. Generator Load Rejection
Turbine Trip No 3.
* N/A
Turbine Trip with Failure of Generator Breakers to Open N/A 4.
: 2. Turbine Trip                               **      No
MSIV Closures No 5.
: 3. Turbine Trip with Failure of Generator Breakers to Open
Pressure Regulator Failure - Open No 6.
* N/A
Pressure Regulator Failure - Closed No 7.
: 4. MSIV Closures                             **      No
Feedwater Controller Failure -
: 5. Pressure Regulator Failure - Open         **      No
Maximum Demand No 8.
: 6. Pressure Regulator Failure - Closed       **      No
Loss of Feedwater Heating No 9.
: 7. Feedwater Controller Failure -
Shutdown Cooling (RHR) Malfunction -
Maximum Demand                             **      No
Decreasing Temperature No 10.
: 8. Loss of Feedwater Heating                 **      No
Inadvertent HPCI Pump Start No 11.
: 9. Shutdown Cooling (RHR) Malfunction -
Continuous Control Rod Withdrawal During Power Range Operation No Event cannot occur.
Decreasing Temperature                     **      No
Bounded by same event at higher power level per FSAR Chapter 15.
: 10. Inadvertent HPCI Pump Start               **        No
i i
: 11. Continuous Control Rod Withdrawal During Power Range Operation               **        No
I
* Event cannot occur.
        **  Bounded by same event at higher power level per FSAR Chapter 15.
i l
i I
l


      .~   .
.~
: 12. Continuous Rod Withdrawal During Reactor Startup                         **    No
12.
: 13. Control Rod Removal Error During Ra' fueling                             **    No   .
Continuous Rod Withdrawal During Reactor Startup No 13.
: 14. Fuel-Assembly Insertion Error During Refueling                         **    No
Control Rod Removal Error During Ra' fueling No 14.
: 15. Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly Into an Improper Location                                 **    No
Fuel-Assembly Insertion Error During Refueling No 15.
: 16. Inadvertent Loading and Operation of a Fuel Assembly in Improper Location                                 **    No
Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly Into an Improper Location No 16.
: 17. Inadvertent Opening of a Safety / Relief Valve                   **    No
Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.
: 18. Loss of Feedwater Flow                   **    No
Inadvertent Opening of a Safety / Relief Valve No 18.
: 19. Loss of AC Power                         **    Yes
Loss of Feedwater Flow No 19.
: 20. Recirculation Pump Trip                 **    No
Loss of AC Power Yes 20.
: 21. Loss of Condenser Vacuum                 **    No
Recirculation Pump Trip No 21.
: 22. Recirculation Pump Seizure               **    No
Loss of Condenser Vacuum No 22.
: 23. Recirculation Flow Control Failure -
Recirculation Pump Seizure No 23.
Decreasing Flow                         **    No
Recirculation Flow Control Failure -
: 24. Recirculation Flow Control Failure -
Decreasing Flow No 24.
With Increasing Flow                     **    No
Recirculation Flow Control Failure -
: 25. Abnormal Startup of Idle Recirculation Pump                       **    No
With Increasing Flow No 25.
: 26. Core Coolant Temperature Increase       **    No
Abnormal Startup of Idle Recirculation Pump No 26.
: 27. Anticipated Transients Without SCRAM (ATWS)                             **    No
Core Coolant Temperature Increase No 27.
: 28. Cask Drop Accident                       *
Anticipated Transients Without SCRAM (ATWS)
                                                                }h0L
No 28.
: 29. Miscellaneous Small Releases Outside Primary Containment             **    No l
Cask Drop Accident
}h0L 29.
Miscellaneous Small Releases Outside Primary Containment No l
l l
l l


i 1
i,
: 30. Off Design Operational Transient l
1 30.
as a Consequence of Instrument Line Failure                           ** No Mk'inCondenserGasTreatment
Off Design Operational Transient l
                  ~
as a Consequence of Instrument Line Failure No 31.
31.
Mk'inCondenserGasTreatment
System Failure                         ** No
~
: 32. Liquid Radwaste Tank Rupture           ** No
No System Failure No 32.
: 33. Control Rod Drop Accident               ** No
Liquid Radwaste Tank Rupture 33.
: 34. Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)     Yes
Control Rod Drop Accident No 34.
: 35. Pipe Breaks Outside Primary Containment (Steam Line Break Accident) ** Yes
Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)
: 36. Fuel Handling Accident                 ** No
Yes 35.
: 37. Feedwater System Piping Break           ** Yes
Pipe Breaks Outside Primary Containment (Steam Line Break Accident) **
: 38. Failure of Air Ejector Lines           ** No i
Yes 36.
l p- -
Fuel Handling Accident No 37.
Feedwater System Piping Break Yes 38.
Failure of Air Ejector Lines No p-
 
ECCS LOCA EVALUATIONS 10 CFR E 50.46 Limits Core Peak Rod Time to 10 PCT Local Core Wide Avg. Powe r MAPLHCR CFR 5 50.46 (F')
Oxidation Oxidation (1 or rated) ikW/ft1 Limits (min)
(Limit 2200')
flimit 171)
(Limit 111 5.0 1.34 55 2200 6.5 less than 0.9 2.5 0.67 124 2200 8.4 less than 1.0 1.25 0.34 285 2100 9.0 1.0
.5 0.13 700 2000 9.0


ECCS LOCA EVALUATIONS 10 CFR E 50.46 Limits Core                                  Peak Rod      Time to 10          PCT                                                          Local  Core Wide    ,,
==1.0 ASSUMPTIONS==
Avg. Powe r                              MAPLHCR      CFR 5 50.46        (F')                                                      Oxidation flimit 171)
10 CFR 50 Appendix K (Standard FSAR Basis)
Oxidation (Limit 111 (1 or rated)                            ikW/ft1      Limits (min)    (Limit 2200')                                                                      .
5.0                                      1.34            55              2200                                                            6.5  less than 0.9 2.5                                      0.67            124              2200                                                            8.4  less than 1.0 1.25                                      0.34            285              2100                                                            9.0  1.0
.5                                        0.13            700              2000                                                            9.0  1.0 ASSUMPTIONS: 10 CFR 50 Appendix K (Standard FSAR Basis)
Initial Conditions Based on Equivalent Core at Designated Core Average Power i
Initial Conditions Based on Equivalent Core at Designated Core Average Power i
td h
td h
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-tTr ft b
ft b


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4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matt'er of
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matt'er of                     )
)
                                                  )
)
LONG ISLAND LIGHTING COMPANY           )   Docket No. 50-322 (OL)
LONG ISLAND LIGHTING COMPANY
                                                  )
)
(Shoreham Nuclear Power Station,       )
Docket No. 50-322 (OL)
Unit 1)                             )
)
AFFIDAVIT OF DR. GLENN G. SHERWOOD, DR. ATAMBIR S. RAO, AND MR. EUGENE C. ECKERT Glenn G. Sherwood, Atambir S. Rao, and Eugene C. Eckert being duly sworn, depose and state as follows:
(Shoreham Nuclear Power Station,
(1)   My name is Glenn G. Sherwood. I am employed by the General Electric Company as Manager, Safety and Licensing Operation. My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.       I have been employed in this position since 1976.     My responsibilities in-clude supervision of the preparation of licensing submittals for General Electric BWRs, including analyses performed in Chapter 15 of safety analysis reports.       In particular, I have been involved in the supervision of licensing matters for the Shoreham Nuclear Power Station since the initial submittal of the Shoreham Final Safety Analysis Report (FSAR).       In this re-gard, I am familiar with the analyses performed in Chapter'15 of that document. From 1974, when I joined General Electric, i
)
Unit 1)
)
AFFIDAVIT OF DR. GLENN G.
: SHERWOOD, DR. ATAMBIR S.
RAO, AND MR. EUGENE C.
ECKERT Glenn G.
Sherwood, Atambir S. Rao, and Eugene C. Eckert being duly sworn, depose and state as follows:
(1)
My name is Glenn G.
Sherwood.
I am employed by the General Electric Company as Manager, Safety and Licensing Operation.
My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.
I have been employed in this position since 1976.
My responsibilities in-clude supervision of the preparation of licensing submittals for General Electric BWRs, including analyses performed in Chapter 15 of safety analysis reports.
In particular, I have been involved in the supervision of licensing matters for the Shoreham Nuclear Power Station since the initial submittal of the Shoreham Final Safety Analysis Report (FSAR).
In this re-gard, I am familiar with the analyses performed in Chapter'15 of that document.
From 1974, when I joined General Electric,


1 l
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~2-l to 1976, I was the Manager, Program Control Section.
                                        ~2-l         to 1976, I was the Manager, Program Control Section. My T
My T
responsibilities included managing engineering and manufacturing work flow for General Electric's nuclear group.
responsibilities included managing engineering and manufacturing work flow for General Electric's nuclear group.
I have a Bachelor of Science degree in Engineering from the U.S. Naval Academy and a Ph.D. in Engineering from the Univer-sity of Michigan.
I have a Bachelor of Science degree in Engineering from the U.S. Naval Academy and a Ph.D. in Engineering from the Univer-sity of Michigan.
(2)   My name is Atambir S. Rao. I am employed by the General Electric Company as Manager, Plant Safety Systems Engi-neering. My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125. I was appointed to my present position in 1984. My responsibilities include ECCS performance analysis, containment performance response analy-sis, and plant safety performance evaluations, including FSAR safety analyses. I have previously held a number of positions relating to accident and transient analyses since I first joined General Electric in 1973. Earlier responsibilities have included modeling and analyzing the thermal hydraulic behavior of BWR fuel following loss of coolant accidents, assessing the implication of advances in heat transfer, fluid mechanics, thermodynamics and two-phase flow on overall BWR system re-sponse during transients and loss of coolant accidents, devel-oping emergency operator guidelines, and assessing containment i
(2)
thermal hydraulic and radiological response for various         )
My name is Atambir S. Rao.
I am employed by the General Electric Company as Manager, Plant Safety Systems Engi-neering.
My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.
I was appointed to my present position in 1984.
My responsibilities include ECCS performance analysis, containment performance response analy-sis, and plant safety performance evaluations, including FSAR safety analyses.
I have previously held a number of positions relating to accident and transient analyses since I first joined General Electric in 1973.
Earlier responsibilities have included modeling and analyzing the thermal hydraulic behavior of BWR fuel following loss of coolant accidents, assessing the implication of advances in heat transfer, fluid mechanics, thermodynamics and two-phase flow on overall BWR system re-sponse during transients and loss of coolant accidents, devel-oping emergency operator guidelines, and assessing containment thermal hydraulic and radiological response for various
)
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                                                                            \
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s   .
s accidents and transients.
accidents and transients. I have been assigned as Manager, Emergency Core Cooling Systems (ECCS) Engineering (1979-80),
I have been assigned as Manager, Emergency Core Cooling Systems (ECCS) Engineering (1979-80),
and Mankger, Containment and Radiological Engineering (1982-84).' I received a Ph.D and a Masters degree in Mechani-cal Engineering from the University of California, Berkeley, and a Bachelor of Technology in Mechanical Engineering from the Indian Institute of Technology, Kanpur, India.
and Mankger, Containment and Radiological Engineering (1982-84).' I received a Ph.D and a Masters degree in Mechani-cal Engineering from the University of California, Berkeley, and a Bachelor of Technology in Mechanical Engineering from the Indian Institute of Technology, Kanpur, India.
(3)   My name is Eugene C. Eckert. I am employed by the General Electric Company as Manager, Power Transient Per-forming Engineering, a position I have held since 1971. My business address is General Electric Company, 175 Curtner Ave-nue, San Jose, California 95125. I am responsible for estab-lishing the simulation requirements of the computer models needed to perform transient analyses, development of design procedures evaluation of BWR stability, and evaluation and specification of the functional protection systems required for reactor abnormal transient protection. Immediately upon joining General Electric Company in September 1959, I partici-pated in assignments which included large jet engine control design, aircraft nuclear propulsion control analysis, nuclear submarine kinetics and control analysis, and industrial control simulation analysis at GE's Research and Development Center.
(3)
My name is Eugene C.
Eckert.
I am employed by the General Electric Company as Manager, Power Transient Per-forming Engineering, a position I have held since 1971.
My business address is General Electric Company, 175 Curtner Ave-nue, San Jose, California 95125.
I am responsible for estab-lishing the simulation requirements of the computer models needed to perform transient analyses, development of design procedures evaluation of BWR stability, and evaluation and specification of the functional protection systems required for reactor abnormal transient protection.
Immediately upon joining General Electric Company in September 1959, I partici-pated in assignments which included large jet engine control design, aircraft nuclear propulsion control analysis, nuclear submarine kinetics and control analysis, and industrial control simulation analysis at GE's Research and Development Center.
In 1962, I joined General Electric's Nuclear Energy Division to l
In 1962, I joined General Electric's Nuclear Energy Division to l


I i
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. 4 work on Boiling Water Reactor simulation and dynamic analysis.
I 1
I have been responsible for design and licensing documentation of the dynamic analysis for several GE BWRs and have partici-pated in initial startup testing of many of the units.
4 work on Boiling Water Reactor simulation and dynamic analysis.
I re-ceived a Bachelor of Science Degree in Electrical Engineering from Valparaiso University in Indiana in 1958.
I have been responsible for design and licensing documentation   l of the dynamic analysis for several GE BWRs and have partici-pated in initial startup testing of many of the units.     I re-ceived a Bachelor of Science Degree in Electrical Engineering from Valparaiso University in Indiana in 1958. I attended Stanford University under an Oak Ridge Fellowship and received a Master of Science Degree in Engineering Science in August 1959.
I attended Stanford University under an Oak Ridge Fellowship and received a Master of Science Degree in Engineering Science in August 1959.
(4) Chapter 15 of the Shoreham FSAR provides the re-sults of analyses for the spectrum of accident and transient events that must be accommodated by the Shoreham plant to dem-onstrate compliance with the NRC's regulations.     This portion of the safety analysis is performed to evaluate the ability of the plant to operate without undue risk to the health and safe-ty of the public. The Shoreham FSAR was submitted to the NRC Staff for review and has been approved by the Staff in its Safety Evaluation Report for Shoreham (NUREG-0420).
(4)
(5)   At the request of the Long Island Lighting Compa-l i
Chapter 15 of the Shoreham FSAR provides the re-sults of analyses for the spectrum of accident and transient events that must be accommodated by the Shoreham plant to dem-onstrate compliance with the NRC's regulations.
ny, General Electric, in conjunction with cognizant LILCO and Stone & Webster personnel, has reviewed all of the events con-   t cidered in Chapter 15 of the FSAR to determine the effect on public health and safety of the operation of the Shoreham plant l
This portion of the safety analysis is performed to evaluate the ability of the plant to operate without undue risk to the health and safe-ty of the public.
                                                                            \
The Shoreham FSAR was submitted to the NRC Staff for review and has been approved by the Staff in its Safety Evaluation Report for Shoreham (NUREG-0420).
(5)
At the request of the Long Island Lighting Compa-l ny, General Electric, in conjunction with cognizant LILCO and i
Stone & Webster personnel, has reviewed all of the events con-t cidered in Chapter 15 of the FSAR to determine the effect on public health and safety of the operation of the Shoreham plant
\\


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  .                                            during fuel load, criticality testing and low power operations.
during fuel load, criticality testing and low power operations.
Although the FSAR considers all phases of the operation of the plant from fuel load to operation at 100% power, this review I
Although the FSAR considers all phases of the operation of the plant from fuel load to operation at 100% power, this review I
was performed specifically to confirm that operation of the Shoreham plant during low power operation will pose no undue risk to public health and safety. The review of Chapter 15 was divided into three parts: (1) fuel load and precriticality testing (Phase I), (2) cold criticality testing (Phase II), and (3) low power testing up to 5% of rated power (Phases III and IV).1/   The review was based upon the same criteria and bases as the original Chapter 15 analyses. Where assumption of a loss or unavailability of offsite power was required in the original analyses, potential unavailability of the TDI diesel generators was considered in this review.
was performed specifically to confirm that operation of the Shoreham plant during low power operation will pose no undue risk to public health and safety.
(6)   The General Electric review of Chapter 15 con-firms that operation during the phases identified above will l
The review of Chapter 15 was divided into three parts: (1) fuel load and precriticality testing (Phase I), (2) cold criticality testing (Phase II), and (3) low power testing up to 5% of rated power (Phases III and IV).1/
i not result in any undue risk to the public health and safety.
The review was based upon the same criteria and bases as the original Chapter 15 analyses.
In fact, the risk from any Chapter 15 event during both the fuel load and precriticality phase and the cold criticality testing phase is essentially non-existent. The risk to the i
Where assumption of a loss or unavailability of offsite power was required in the original analyses, potential unavailability of the TDI diesel generators was considered in this review.
l l       1/   Parts (1) and (2) correspond to Phases I and II, respec-tively, as described in the Affidavit of Messrs. Notaro and Gunther. Part (3) corresponds to Phases III and IV, combined, as described in that affidavit.
(6)
The General Electric review of Chapter 15 con-firms that operation during the phases identified above will l
not result in any undue risk to the public health and safety.
i In fact, the risk from any Chapter 15 event during both the fuel load and precriticality phase and the cold criticality testing phase is essentially non-existent.
The risk to the i
l l
1/
Parts (1) and (2) correspond to Phases I and II, respec-tively, as described in the Affidavit of Messrs. Notaro and Gunther.
Part (3) corresponds to Phases III and IV, combined, as described in that affidavit.


        -                                                                    1 e   ,
1 e
i public health and safety from the Chapter 15 events postulated for low power testing up to 5% of rated power is small in com-parison I to the risks already found acceptable for 100% power operation.' As already indicated, this review considered the impact of potential diesel unavailability.
i,
Phase I:   Fuel Loading and Precriticality Testing (7)   This phase of operation of the Shoreham plant in-cludes only initial fuel loading and precriticality testing.
public health and safety from the Chapter 15 events postulated for low power testing up to 5% of rated power is small in com-I parison to the risks already found acceptable for 100% power operation.' As already indicated, this review considered the impact of potential diesel unavailability.
The reactor will remain at essentially ambient temperature and
Phase I:
,          atmospheric pressure. The reactor will not be taken critical.
Fuel Loading and Precriticality Testing (7)
Any increase in temperature beyond ambient conditions will be due only to external heat sources such as recirculation pump heat. There will be no heat generation in the core. Details of the steps to be performed during these operations are de-scribed in the Phase I discussion in the affidavit submitted by Messrs. Notaro and Gunther.
This phase of operation of the Shoreham plant in-cludes only initial fuel loading and precriticality testing.
(8)   The review of the Chapter 15 analysis revealed that of the 38 accident or transient events addressed in Chap-ter 15, 18 of the events could not occur during Phase I because of the operating conditions of the plant. An additional 5 events could physically occur, but given the plant conditions, could not constitute events in the context of the Chapter 15     l l
The reactor will remain at essentially ambient temperature and atmospheric pressure.
The reactor will not be taken critical.
Any increase in temperature beyond ambient conditions will be due only to external heat sources such as recirculation pump heat.
There will be no heat generation in the core.
Details of the steps to be performed during these operations are de-scribed in the Phase I discussion in the affidavit submitted by Messrs. Notaro and Gunther.
(8)
The review of the Chapter 15 analysis revealed that of the 38 accident or transient events addressed in Chap-ter 15, 18 of the events could not occur during Phase I because of the operating conditions of the plant.
An additional 5 events could physically occur, but given the plant conditions, could not constitute events in the context of the Chapter 15 l


,  s   .
s 1
1 3
3 i
i i                                  safety analysis. The remaining 15 events could possibly occur, 2
i safety analysis.
although occurrence is highly unlikely given the plant condi-2 tions. EIn   any event, it is readily apparent that the potential i                                   consequences of these 15 events would be trivial. Exhibit 1 l                                   below lists the category into which each Chapter 15 event l                                   falls.
The remaining 15 events could possibly occur, 2
j                                           (9)   The 18 Chapter 15 events which could not occur   l
although occurrence is highly unlikely given the plant condi-2 tions. EIn any event, it is readily apparent that the potential i
!                                                                                                    l during Phase I are precluded by the operating conditions of the 4
consequences of these 15 events would be trivial.
reactor. These events all involve operating modes or component operation which are not possible during this phase. For exam-t ple, during fuel loading and precriticality testing, the reac-tor is at essentially ambient temperature and atmospheric pres-sure. Accordingly, no steam is available. Thus, all events which would require pressurized conditions are precluded.
Exhibit 1 l
below lists the category into which each Chapter 15 event l
falls.
j (9)
The 18 Chapter 15 events which could not occur l
l during Phase I are precluded by the operating conditions of the 4
reactor.
These events all involve operating modes or component operation which are not possible during this phase.
For exam-t ple, during fuel loading and precriticality testing, the reac-tor is at essentially ambient temperature and atmospheric pres-sure.
Accordingly, no steam is available.
Thus, all events which would require pressurized conditions are precluded.
Events such as turbine trip (FSAR $ 15A.1.2), loss of feedwater l
Events such as turbine trip (FSAR $ 15A.1.2), loss of feedwater l
heating (FSAR S 15A.1.8) and inadvertent opening of a safety
heating (FSAR S 15A.1.8) and inadvertent opening of a safety relief valve require the generation of steam for the event to l
!                                  relief valve require the generation of steam for the event to l
occur.
!                                  occur. Similarly, there is no steam flow to interrupt, thus i
Similarly, there is no steam flow to interrupt, thus i
precluding an MSIV closure event (FSAR S 15A.1.4). Other events are precluded by definition. Thus, events such as con-l l                                   tinuous control rod withdrawal during power range operation l
precluding an MSIV closure event (FSAR S 15A.1.4).
Other events are precluded by definition.
Thus, events such as con-l l
tinuous control rod withdrawal during power range operation l
(FSAR 5 15A.1.11) and operation of a fuel assembly in an im-proper location (FSAR $ 15A.1.16) cannot be postulated.
(FSAR 5 15A.1.11) and operation of a fuel assembly in an im-proper location (FSAR $ 15A.1.16) cannot be postulated.


s   .
s (10)
1
In addition to the 18 events which simply cannot occur, there are 5 events for which the component operation evaluate'd in Chapter 15 could occur, but the phenomena of in-terest in Chapter 15 could not exist.
  .                                                            (10) In addition to the 18 events which simply cannot occur, there are 5 events for which the component operation evaluate'd in Chapter 15 could occur, but the phenomena of in-terest in Chapter 15 could not exist. All recirculation pump events, such as recirculation pump trip (FSAR $ 15A.1.20) and abnormal startup of an idle recirculation pump (FSAR S 15A.1.25), would be of interest only if they could affect core physics or thermal-hydraulic conditions. With no heat generation or boiling in the core, there are no pertinent phe-nomena (such as temperature differences or void collapses) to evaluate. Another example, the core coolant temperature in-crease event (FSAR $ 15.A.1.26), postulates a loss of RHR cool-ing. Even if the RHR system was operated in Phase I, there would be no temperature increase from decay heat to evaluate should the RHR system be lost.
All recirculation pump events, such as recirculation pump trip (FSAR $ 15A.1.20) and abnormal startup of an idle recirculation pump (FSAR S 15A.1.25), would be of interest only if they could affect core physics or thermal-hydraulic conditions.
(11) The remaining 15 events addressed in Chapter 15 could possibly occur. However, our review established that all are trivial events which have no potential to impact public health and safety. Prior to initial criticality,.there are no fission products in the core and no decay heat exists. It fol-lows that core cooling is not required. In addition, with no fission product inventory, there are no fission product re-l leases possible. Thus, for reactor events such as a control l
With no heat generation or boiling in the core, there are no pertinent phe-nomena (such as temperature differences or void collapses) to evaluate.
Another example, the core coolant temperature in-crease event (FSAR $ 15.A.1.26), postulates a loss of RHR cool-ing.
Even if the RHR system was operated in Phase I, there would be no temperature increase from decay heat to evaluate should the RHR system be lost.
(11)
The remaining 15 events addressed in Chapter 15 could possibly occur.
However, our review established that all are trivial events which have no potential to impact public health and safety.
Prior to initial criticality,.there are no fission products in the core and no decay heat exists.
It fol-lows that core cooling is not required.
In addition, with no fission product inventory, there are no fission product re-l leases possible.
Thus, for reactor events such as a control l
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1
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's
      -                                                                    1
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_g_
rod removal error (FSAR 5 15A.1.13) and a control rod drop (FSAR S 15.1.33) and for non-reactor events such as a fuel han-dling abeident (FSAR 5 15.1.36) or a liquid radwaste tank rup-ture (FSAR~5 15.1.32), there could be no radiological conse-quences. Therefore, there is no risk to public health and safety.
rod removal error (FSAR 5 15A.1.13) and a control rod drop (FSAR S 15.1.33) and for non-reactor events such as a fuel han-dling abeident (FSAR 5 15.1.36) or a liquid radwaste tank rup-ture (FSAR~5 15.1.32), there could be no radiological conse-quences.
(12) Even a loss of coolant accident (FSAR 5 15.1.34) could have no radiological consequences during Phase I. No core cooling is required. No fission product release is possi-ble. The fuel simply could not be challenged by a complete draindown of the reactor vessel for an unlimited period of time.
Therefore, there is no risk to public health and safety.
(13) In summary, the review of Chapter 15 events for fuel loading and precriticality testing indicates that many Chapter 15 events simply cannot occur, and for those that can, there can be no radiological consequences. Therefore, there is no possible risk to the public health and safety. This conclu-sion is not affected by any postulated diesel generator unavailability because it is in no way dependent on the avail-ability or unavailability of any AC power.
(12)
Phase II:   Cold Criticality Testing (14) This phase of low power testing of the Shoreham
Even a loss of coolant accident (FSAR 5 15.1.34) could have no radiological consequences during Phase I.
No core cooling is required.
No fission product release is possi-ble.
The fuel simply could not be challenged by a complete draindown of the reactor vessel for an unlimited period of time.
(13)
In summary, the review of Chapter 15 events for fuel loading and precriticality testing indicates that many Chapter 15 events simply cannot occur, and for those that can, there can be no radiological consequences.
Therefore, there is no possible risk to the public health and safety.
This conclu-sion is not affected by any postulated diesel generator unavailability because it is in no way dependent on the avail-ability or unavailability of any AC power.
Phase II:
Cold Criticality Testing (14)
This phase of low power testing of the Shoreham


plant will include cold criticality testing of the plant at es-sentially ambient temperature and atmospheric pressure. The power level during this phase of testing will be in the range of .0001% to .001% of rated power. Details of the testing to be performed during this phase are described in the Notaro Af-fidavit.
plant will include cold criticality testing of the plant at es-sentially ambient temperature and atmospheric pressure.
(15) The review of Chapter 15 revealed that of the 38 accident or transient events included there, 15 of the events could not occur because of the operating conditions of the plant during Phase I. See Exhibit 2. A number of these events are not possible because the reactor will be at essentially ambient temperature and pressure and no steam will be gener-ated. For example, the generator load rejection event (FSAR S 15A.1.1) could not occur during this testing phase because steam is needed to drive the main turbine generator to permit ccnnecting it to the LILCO transmission system. Another exam-ple, the loss of condenser vacuum event (FSAR 5 15A.1.21),
The power level during this phase of testing will be in the range of.0001% to.001% of rated power.
could not occur because it assumes that steam is available to draw a vacuum in the main condenser. A third example, the in-advertent HPCI pump start event (FSAR S 15A.1.10), could not occur because there will be no steam available to power the HPCI pump, a steam driven ECCS pump. Other Chapter 15 events could not occur because they are precluded by the configuration I
Details of the testing to be performed during this phase are described in the Notaro Af-fidavit.
i 1
(15)
The review of Chapter 15 revealed that of the 38 accident or transient events included there, 15 of the events could not occur because of the operating conditions of the plant during Phase I.
See Exhibit 2.
A number of these events are not possible because the reactor will be at essentially ambient temperature and pressure and no steam will be gener-ated.
For example, the generator load rejection event (FSAR S 15A.1.1) could not occur during this testing phase because steam is needed to drive the main turbine generator to permit ccnnecting it to the LILCO transmission system.
Another exam-ple, the loss of condenser vacuum event (FSAR 5 15A.1.21),
could not occur because it assumes that steam is available to draw a vacuum in the main condenser.
A third example, the in-advertent HPCI pump start event (FSAR S 15A.1.10), could not occur because there will be no steam available to power the HPCI pump, a steam driven ECCS pump.
Other Chapter 15 events could not occur because they are precluded by the configuration I
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l
.i l .                                                                       l of the plant during this phase of low power testing. An exam-ple of this type of event is the MSIV closure (FSAR $ 15A.1.4).
.i l.
The MSIds will normally be closed throughout all of the op-erations conducted during this phase of low power testing.     In any event, there is no steam generated by the reactor to flow through the steam lines.
of the plant during this phase of low power testing.
(16)   In addition to the 15 events that could not occur during Phase I, many of the 23 events remaining in the Chapter 15 analysis are far less likely to occur during low power testing than during normal operations. For example, the recirculation pump trip (FSAR 9 15A.1.20), the recirculation pump seizure (FSAR S 15A.1.22), the recirculation flow control failures (FSAR $ 15A.1.23 and 24) and the abnormal startup of idle recirculation pump (FSAR $ 15A.1.25) events, although physically possible, are not as likely to occur because the recirculation pumps are used for only limited periods of time during this phase of the testing program. Similarly, the loss of feedwater event (FSAR $ 15A.1.18) is very unlikely because little, if any, make-up water will have to be supplied to the reactor. Moreover, make-up water would not normally be supplied by the feedwater system under these conditions. Other-very unlikely events include miscellaneous small releases out-side primary containment (FSAR 5 15.1.29), off design
An exam-ple of this type of event is the MSIV closure (FSAR $ 15A.1.4).
The MSIds will normally be closed throughout all of the op-erations conducted during this phase of low power testing.
In any event, there is no steam generated by the reactor to flow through the steam lines.
(16)
In addition to the 15 events that could not occur during Phase I, many of the 23 events remaining in the Chapter 15 analysis are far less likely to occur during low power testing than during normal operations.
For example, the recirculation pump trip (FSAR 9 15A.1.20), the recirculation pump seizure (FSAR S 15A.1.22), the recirculation flow control failures (FSAR $ 15A.1.23 and 24) and the abnormal startup of idle recirculation pump (FSAR $ 15A.1.25) events, although physically possible, are not as likely to occur because the recirculation pumps are used for only limited periods of time during this phase of the testing program.
Similarly, the loss of feedwater event (FSAR $ 15A.1.18) is very unlikely because little, if any, make-up water will have to be supplied to the reactor.
Moreover, make-up water would not normally be supplied by the feedwater system under these conditions.
Other-very unlikely events include miscellaneous small releases out-side primary containment (FSAR 5 15.1.29), off design


t i
t i '
operational transient as a consequence of instrument line fail-ure (FSAR $ 15.1.30), and feedwater system piping break (FSAR
operational transient as a consequence of instrument line fail-ure (FSAR $ 15.1.30), and feedwater system piping break (FSAR
                  $ 15.1.37). Thus, many of the Chapter 15 events that are phys-4                ically possible during Phase II remain very unlikely in light of the plant conditions that will then exist.
$ 15.1.37).
I (17) Nonetheless, all 23 possible events contained in Chapter 15 were reviewed to reaffirm that the consequences of these events, should one occur during Phase I of low power testing, would be bounded by the consequences analyzed for the
Thus, many of the Chapter 15 events that are phys-ically possible during Phase II remain very unlikely in light 4
;                event considered in the FSAR.                   A discussion of some of the 23 l
of the plant conditions that will then exist.
l                 possible events contained in Chapter 15 illustrates the basis for this conclusion.               The continuous control rod withdrawal
I (17)
,                during startup event (FSAR $ 15A.1.12) is applicable to op-eration in the power, source and/or intermediate range of op-eration. During cold functional criticality testing, the reac-tor will operate in the source and intermediate ranges and therefore the conclusions contained in Chapter 15 are applica-j ble to this event should it occur during this phase of low power testing.       As the-FSAR indicates, this event would not re-sult in any release of radioactive material from the fuel at any power level.       -Another example is the fuel handling accident j                 (FSAR $ 15.1.36).       As stated in the FSAR, the_most severe fuel handling accident from a radiological viewpoint is a dropping b
Nonetheless, all 23 possible events contained in Chapter 15 were reviewed to reaffirm that the consequences of these events, should one occur during Phase I of low power testing, would be bounded by the consequences analyzed for the event considered in the FSAR.
            - - -        -        ---        .,-ee   g .!.--            ,    y       ~- _ . _ - - . _ . . - . - - - -    %
A discussion of some of the 23 l
l possible events contained in Chapter 15 illustrates the basis for this conclusion.
The continuous control rod withdrawal during startup event (FSAR $ 15A.1.12) is applicable to op-eration in the power, source and/or intermediate range of op-eration.
During cold functional criticality testing, the reac-tor will operate in the source and intermediate ranges and therefore the conclusions contained in Chapter 15 are applica-j ble to this event should it occur during this phase of low power testing.
As the-FSAR indicates, this event would not re-sult in any release of radioactive material from the fuel at any power level.
-Another example is the fuel handling accident j
(FSAR $ 15.1.36).
As stated in the FSAR, the_most severe fuel handling accident from a radiological viewpoint is a dropping b
.,-ee g
y
~-


of the fuel assembly onto the top of the core. The ESAR analy-sis assumes that the fuel contains a fission product inventory equivaleht to operation of 1000 days at full rated power. This -
of the fuel assembly onto the top of the core.
assumption results in an equilibrium fission product concentra-tion at the time the reactor is shut down. But as already noted, the fission product inventories in the core will be sig-nificantly less during Phase II low power testing than the in-ventories analyzed in the FSAR because of the extremely low power levels (.0001% to .001% of rated power) achieved during this testing. Thus, even if a handling accident took place and fuel damage did occur, there would be significantly less fis-sion products to be released from the core than those that have already been analyzed and found accept'able in the FSAR. A third example is the liquid radwaste tank rupture event (FSAR $
The ESAR analy-sis assumes that the fuel contains a fission product inventory equivaleht to operation of 1000 days at full rated power.
15.1.32). This event assumes the rupture of a radwaste tank that contains a substantial amount of contaminated liquids gen-erated during the operation of the reactor. But again, since Phase II low power testing results in insignificant power lev-els in the reactor, there will be little, if any, radioactive liquids in the radwaste tank should such a rupture occur.
This assumption results in an equilibrium fission product concentra-tion at the time the reactor is shut down.
Thus, even the minimal consequences already described in the FSAR for the design basis event would be further reduced under these low power testing conditions. For each of these events, the review concluded that the consequences are significantly
But as already noted, the fission product inventories in the core will be sig-nificantly less during Phase II low power testing than the in-ventories analyzed in the FSAR because of the extremely low power levels (.0001% to.001% of rated power) achieved during this testing.
Thus, even if a handling accident took place and fuel damage did occur, there would be significantly less fis-sion products to be released from the core than those that have already been analyzed and found accept'able in the FSAR.
A third example is the liquid radwaste tank rupture event (FSAR $
15.1.32).
This event assumes the rupture of a radwaste tank that contains a substantial amount of contaminated liquids gen-erated during the operation of the reactor.
But again, since Phase II low power testing results in insignificant power lev-els in the reactor, there will be little, if any, radioactive liquids in the radwaste tank should such a rupture occur.
Thus, even the minimal consequences already described in the FSAR for the design basis event would be further reduced under these low power testing conditions.
For each of these events, the review concluded that the consequences are significantly


, ,                                                          less severe for any event occurring during the cold functional   l criticality testing than for the event analyzed in Chapter 15.
less severe for any event occurring during the cold functional criticality testing than for the event analyzed in Chapter 15.
To summatize, because of the extremely low power levels reached during this' testing phase, fission product inventory in the core will be only a small fraction of that assumed for the Chapter 15 analyses. As indicated above, the FSAR assumes op-eration at 100% power for 1000 days in calculating fission product inventory; the inventory during Phase II low power testing will be less than one one-hundred-thousandth (.00001) of the fission product inventory assumed in the FSAR. Conse-quently, none of the events analyzed in Chapter 15 could result in a release of radioactivity during cold criticality testing that would harm the public health and safety.
To summatize, because of the extremely low power levels reached during this' testing phase, fission product inventory in the core will be only a small fraction of that assumed for the Chapter 15 analyses.
(18) The review of Chapter 15 events for Phase II testing and the conclusions reached are unaffected by any unavailability of the TDI diesels. Of the 23 possible Chapter 15 events reviewed, 20 of the events in the FSAR do not require the assumption of loss or unavailability of offsite AC power.
As indicated above, the FSAR assumes op-eration at 100% power for 1000 days in calculating fission product inventory; the inventory during Phase II low power testing will be less than one one-hundred-thousandth (.00001) of the fission product inventory assumed in the FSAR.
See Exhibit 2. Thus, our conclusions for these 20 of the 23 possible events are independent of the status of the diesels.
Conse-quently, none of the events analyzed in Chapter 15 could result in a release of radioactivity during cold criticality testing that would harm the public health and safety.
(18)
The review of Chapter 15 events for Phase II testing and the conclusions reached are unaffected by any unavailability of the TDI diesels.
Of the 23 possible Chapter 15 events reviewed, 20 of the events in the FSAR do not require the assumption of loss or unavailability of offsite AC power.
See Exhibit 2.
Thus, our conclusions for these 20 of the 23 possible events are independent of the status of the diesels.
I I
I I
(19) The three events'that do assume loss or unavailability of offsite AC power are (1) pipe breaks inside the primary containment (LOCA) (FSAR $ 15.1.34), (2) feedwater l
(19)
The three events'that do assume loss or unavailability of offsite AC power are (1) pipe breaks inside the primary containment (LOCA) (FSAR $ 15.1.34), (2) feedwater l
l l
l l


system piping break (FSAR $ 15.1.37), and (3) the loss of AC power event (FSAR S 15A.1.19). With respect to these events, the LOCI'would be the most limiting event. The review has shown that'if a LOCA did occur during the cold criticality testing phase, however remote that possibility, there would be time on the order of months available to restore make-up water for core cooling. At the power levels achieved during Phase II, fission product inventory is very low. At most, decay heat ,
system piping break (FSAR $ 15.1.37), and (3) the loss of AC power event (FSAR S 15A.1.19).
will, on the average, be a fraction of a watt per rod, with no single rod exceeding approximately 2 watts. This is less, roughly, than the heat output of a Christmas tree bulb.       It follows that the fuel cladding temperature would not exceed the limits of 10 CFR $ 50.46 and Appendix K even after months with-out cooling and without any source of AC power.
With respect to these events, the LOCI'would be the most limiting event.
(20)   The loss of AC power event (FSAR S 15A.1.19) and the feedwater system piping break (FSAR S 15.1.37) under cold criticality testing conditions do not rely on the diesel gener-ators for mitigation of the event.     For these events, since no loss of coolant occurs and the decay heat is minimal, core cooling is achieved, without AC power, using the existing core water inventory and heat losses to ambient, for essentially un-limited periods of time.     In'any event, as demonstrated in the l       Schiffmacher Affidavit, AC power sources can and will be l
The review has shown that'if a LOCA did occur during the cold criticality testing phase, however remote that possibility, there would be time on the order of months available to restore make-up water for core cooling.
l l
At the power levels achieved during Phase II, fission product inventory is very low.
i I
At most, decay heat will, on the average, be a fraction of a watt per rod, with no single rod exceeding approximately 2 watts.
This is less, roughly, than the heat output of a Christmas tree bulb.
It follows that the fuel cladding temperature would not exceed the limits of 10 CFR $ 50.46 and Appendix K even after months with-out cooling and without any source of AC power.
(20)
The loss of AC power event (FSAR S 15A.1.19) and the feedwater system piping break (FSAR S 15.1.37) under cold criticality testing conditions do not rely on the diesel gener-ators for mitigation of the event.
For these events, since no loss of coolant occurs and the decay heat is minimal, core cooling is achieved, without AC power, using the existing core water inventory and heat losses to ambient, for essentially un-limited periods of time.
In'any event, as demonstrated in the l
Schiffmacher Affidavit, AC power sources can and will be i
I


readily supplied to the Shoreham plant even if one assumes the     .
readily supplied to the Shoreham plant even if one assumes the simultaneous loss of all three emergency diesel generators.
1 simultaneous loss of all three emergency diesel generators.
(21)
(21) In addition to our conclusions that the limiting LOCA event could not approach the limits of 10 CFR 6 50.46 and     l Appendix K during Phase II low power testing, there are other reasons why our findings with respect to the three events that assume loss of AC power are independent of the availability of the TDI diesels. The LOCA (pipe break inside containment) and the feedwater system piping break postulate the double ended rupture of a piping system. Because the reactor will be at es-sentially ambient temperature and atmospheric pressure during Phase II, it is extremely unlikely that such a pipe break would ever occur. In fact, the NRC Staff does not require double ended ruptures to be postulated for low temperature and low pressure systems in safety analyses. Thus, these events are much less likely during cold criticality testing than during normal operation.
In addition to our conclusions that the limiting LOCA event could not approach the limits of 10 CFR 6 50.46 and Appendix K during Phase II low power testing, there are other reasons why our findings with respect to the three events that assume loss of AC power are independent of the availability of the TDI diesels.
i (22) The review of Chapter 15 events for cold criti-cality testing indicates that performance of these activities at Shoreham involves essentially no risk to the public health and safety. This conclusion is not affected by any postulated r
The LOCA (pipe break inside containment) and the feedwater system piping break postulate the double ended rupture of a piping system.
diesel unavailability. In fact, even if AC power were not available for extended periods of time, fuel design limits and I
Because the reactor will be at es-sentially ambient temperature and atmospheric pressure during Phase II, it is extremely unlikely that such a pipe break would ever occur.
_  .                              -  ~
In fact, the NRC Staff does not require double ended ruptures to be postulated for low temperature and low pressure systems in safety analyses.
Thus, these events are much less likely during cold criticality testing than during normal operation.
i (22)
The review of Chapter 15 events for cold criti-cality testing indicates that performance of these activities at Shoreham involves essentially no risk to the public health and safety.
This conclusion is not affected by any postulated r
diesel unavailability.
In fact, even if AC power were not available for extended periods of time, fuel design limits and I
~


l                                                                                                      l 17-design conditions of the reactor coolant pressure boundary would not be approached or exceeded as a result of anticipated operatihnaloccurrences, and the core would be adequately cooled in the unlikely event of a postulated accident.
l 17-design conditions of the reactor coolant pressure boundary would not be approached or exceeded as a result of anticipated operatihnaloccurrences, and the core would be adequately cooled in the unlikely event of a postulated accident.
Phases III and IV:
Phases III and IV:
Low Power Testing Up To 5% of Rated Power (23)   These aspects of low power testing will include operation of the plant at power levels up to 5% of rated power.
Low Power Testing Up To 5% of Rated Power (23)
These aspects of low power testing will include operation of the plant at power levels up to 5% of rated power.
Details of the testing to be performed during this phase of op-eration are described as Phases III and IV in the Notaro Affi-davit.
Details of the testing to be performed during this phase of op-eration are described as Phases III and IV in the Notaro Affi-davit.
(24) The review of the 38 Chapter 15 events for these phases of low power testing operations revealed that two of the events in Chapter 15, generator load rejection-(FSAR 6 15A.1.1) and turbine trip with generator breaker failure (FSAR 6 15.1.2) cannot occur because the generator will not be connected to the-grid during these phases of testing.     A third event,.the cask drop, is precluded by desiga as stated in FSAR 5 15.1.28.                   See Exhibit 3.
(24)
(25) of the remaining 35 events that can occur during this phase of operation, 31 of the events do not assume loss or 4
The review of the 38 Chapter 15 events for these phases of low power testing operations revealed that two of the events in Chapter 15, generator load rejection-(FSAR 6 15A.1.1) and turbine trip with generator breaker failure (FSAR 6 15.1.2) cannot occur because the generator will not be connected to the-grid during these phases of testing.
unavailability of AC power.     For each of these 31 events, 4
A third event,.the cask drop, is precluded by desiga as stated in FSAR 5 15.1.28.
                                                                                              ~
See Exhibit 3.
w r --< - - - < - v a -r--e - - =- - ~   t -w -f
(25) of the remaining 35 events that can occur during this phase of operation, 31 of the events do not assume loss or 4
unavailability of AC power.
For each of these 31 events, 4
t w
r v a
-r--e
- - =- - ~
~
t -w -f


                    ~
~.
      ,  operation of the plant up to 5% of rated power will be bounded by the Chapter 15 analysis. Since the Chapter 15 analysis con-siders all possible phases of plant operation, it follows that operation at 5% can result in consequences less severe than those analyzed in Chapter 15. For example, the turbine trip event (FSAR $ 15A.1.2) assumes that the limiting event occurs with the reactor operating at 105% of rated steam flow coupled with failure of the turbine bypass valves to open. Even this limiting event does not result in any fuel failures. FSAR S 15A.1.2 specifically notes that turbine trips at power levels
operation of the plant up to 5% of rated power will be bounded by the Chapter 15 analysis.
,          less than 30% of rated power are bounded by the limiting analy-sis. Another example is the loss of feedwater heating event (FSAR S 15A.1.8). This event assumes continuous operation of the feedwater system and the most severe possible loss of feedwater heating, resulting in the injection of colder feedwater. For operation at power levels less than 5%, the im-pact of lost feedwater heating is minimal because of the low feedwater flow. Since these analyses are not required to as-sume the absence of AC power, potential unavailability of the TDI diesels has no effect on the assessment of these events.
Since the Chapter 15 analysis con-siders all possible phases of plant operation, it follows that operation at 5% can result in consequences less severe than those analyzed in Chapter 15.
(26)   Not only are the results of these 31 events bounded by the Chapter 15 analysis, the consequences of these events are also less than the consequences stated in the FSAR.
For example, the turbine trip event (FSAR $ 15A.1.2) assumes that the limiting event occurs with the reactor operating at 105% of rated steam flow coupled with failure of the turbine bypass valves to open.
Even this limiting event does not result in any fuel failures.
FSAR S 15A.1.2 specifically notes that turbine trips at power levels less than 30% of rated power are bounded by the limiting analy-sis.
Another example is the loss of feedwater heating event (FSAR S 15A.1.8).
This event assumes continuous operation of the feedwater system and the most severe possible loss of feedwater heating, resulting in the injection of colder feedwater.
For operation at power levels less than 5%, the im-pact of lost feedwater heating is minimal because of the low feedwater flow.
Since these analyses are not required to as-sume the absence of AC power, potential unavailability of the TDI diesels has no effect on the assessment of these events.
(26)
Not only are the results of these 31 events bounded by the Chapter 15 analysis, the consequences of these events are also less than the consequences stated in the FSAR.
l
l


First, the power limitations during low power testing up to 5%
First, the power limitations during low power testing up to 5%
power, the fission product inventory in the core will not ex-ceed 5%b of the values assumed in the FSAR. In fact, because of the intermittant type of operations conducted during low power testing, equilibrium fission product inventory for even 5%
power, the fission product inventory in the core will not ex-b ceed 5% of the values assumed in the FSAR.
power is unlikely to be achieved. This low fission product in-ventory reduces risk in two ways:   (a) the amount of decay heat present in the core following shutdown is substantially re-duced, and (b) the amount of radioactivity that could be re-leased upon fuel failure is substantially reduced.
In fact, because of the intermittant type of operations conducted during low power testing, equilibrium fission product inventory for even 5%
(27)   The second factor contributing to the signifi-cantly lower risk during low power operation is the increased time available for preventive or mitigating action should such action be deemed desirable by the operator. Longer time is
power is unlikely to be achieved.
!          available because the limited power levels mean that it takes longer for the plant to reach setpoints and limits. For exam-ple, on loss of feedwater (FSAR S 15A.1.18), the water level in the reactor will decrease at a slower rate than if the event occured at 100% power. This gives the operator more time to act manually to restore feedwater before an automatic action takes place. Similarly, in the loss of condenser vacuum event (FSAR 5 15.A.1.21), the operator will have more time to identi-fy the decreasing vacuum and to take steps to remedy the
This low fission product in-ventory reduces risk in two ways:
(a) the amount of decay heat present in the core following shutdown is substantially re-duced, and (b) the amount of radioactivity that could be re-leased upon fuel failure is substantially reduced.
(27)
The second factor contributing to the signifi-cantly lower risk during low power operation is the increased time available for preventive or mitigating action should such action be deemed desirable by the operator.
Longer time is available because the limited power levels mean that it takes longer for the plant to reach setpoints and limits.
For exam-ple, on loss of feedwater (FSAR S 15A.1.18), the water level in the reactor will decrease at a slower rate than if the event occured at 100% power.
This gives the operator more time to act manually to restore feedwater before an automatic action takes place.
Similarly, in the loss of condenser vacuum event (FSAR 5 15.A.1.21), the operator will have more time to identi-fy the decreasing vacuum and to take steps to remedy the


situation before automatic actions such as turbine trip, feedpump trip or main steam isolation occur. Another example is the mhin steam isolation valve closure event (FSAR S 15A.l.4).' At five percent power, the amount of heat produced upon isolation of the reactor vessel (which is followed by a reactor scram) results in a much slower pressure and tempera-ture increase than would be experienced at 100% power. This gives the operator more time to manually initiate reactor cool-ing rather than relying on automatic action. In effect, the operator may end the transient before there is any substantial impact on the plant.
situation before automatic actions such as turbine trip, feedpump trip or main steam isolation occur.
(28) The third factor contributing to the signifi-cantly lower risk during low power testing is the reduction in the required capacity for mitigating systems. Because of the lower levels of decay heat present following operation at 5%
Another example is the mhin steam isolation valve closure event (FSAR S 15A.l.4).'
power, the demand for core cooling and auxiliary systems is substantially reduced, permitting the operation of fewer sys-tems and components to mitigate any event. It follows that the AC power requirements for event mitigation are substantially reduced for 5% power operation as compared to 100% power op-eration.
At five percent power, the amount of heat produced upon isolation of the reactor vessel (which is followed by a reactor scram) results in a much slower pressure and tempera-ture increase than would be experienced at 100% power.
(29) As already noted, only four of the events ana-lyzed in Chapter 15 require the assumption of the 1
This gives the operator more time to manually initiate reactor cool-ing rather than relying on automatic action.
In effect, the operator may end the transient before there is any substantial impact on the plant.
(28)
The third factor contributing to the signifi-cantly lower risk during low power testing is the reduction in the required capacity for mitigating systems.
Because of the lower levels of decay heat present following operation at 5%
power, the demand for core cooling and auxiliary systems is substantially reduced, permitting the operation of fewer sys-tems and components to mitigate any event.
It follows that the AC power requirements for event mitigation are substantially reduced for 5% power operation as compared to 100% power op-eration.
(29)
As already noted, only four of the events ana-lyzed in Chapter 15 require the assumption of the 1
t
t


unavailability of offsite AC power for operation during Phases III and IV. Of these four events, the loss of coolant accident is the     ost limiting event. The Chapter 15 LOCA analysis as-sumes the unavailability of offsite AC power.     This is a con-servative licensing assumption.     In fact, as described in de-tail in the Schiffmacher Affidavit, there are multiple sources of AC power available to the Shoreham site (e.g., emergency diesel generators, two normal sources of offsite power, blackstart gas turbines at Holtsville, Southhold, and East Hampton, a blackstart gas turbine on the Shoreham site, and mo-bile diesel generators). Thus, AC power will be available at Shoreham to mitigate a loss of coolant accident during low power operations up to 5% rated power.       In the unlikely event offsite AC power is lost, it can be restored within sufficient time to prevent exceeding the limits of 10 CFR $ 50.46 and Ap-pendix K. GE has determined that for 5% power so long as reflooding of the core has occurred within approximately one hour, S 50.46 criteria will be met.2/ As the Schiffmacher Af-fidavit demonstrates, power can be restored to Shoreham within minutes. An evaluation has been performed to assure the ade-quacy of containment isolation in the event AC power sources 2/     As shown in the Exhibit 4 below, lower power levels will result in more time to restore power and core cooling for a postulated LOCA. Thus, for 1% power approximately 5 hours are available.
unavailability of offsite AC power for operation during Phases III and IV.
Of these four events, the loss of coolant accident is the ost limiting event.
The Chapter 15 LOCA analysis as-sumes the unavailability of offsite AC power.
This is a con-servative licensing assumption.
In fact, as described in de-tail in the Schiffmacher Affidavit, there are multiple sources of AC power available to the Shoreham site (e.g., emergency diesel generators, two normal sources of offsite power, blackstart gas turbines at Holtsville, Southhold, and East Hampton, a blackstart gas turbine on the Shoreham site, and mo-bile diesel generators).
Thus, AC power will be available at Shoreham to mitigate a loss of coolant accident during low power operations up to 5% rated power.
In the unlikely event offsite AC power is lost, it can be restored within sufficient time to prevent exceeding the limits of 10 CFR $ 50.46 and Ap-pendix K.
GE has determined that for 5% power so long as reflooding of the core has occurred within approximately one hour, S 50.46 criteria will be met.2/ As the Schiffmacher Af-fidavit demonstrates, power can be restored to Shoreham within minutes.
An evaluation has been performed to assure the ade-quacy of containment isolation in the event AC power sources 2/
As shown in the Exhibit 4 below, lower power levels will result in more time to restore power and core cooling for a postulated LOCA.
Thus, for 1% power approximately 5 hours are available.
t i
t i
l
l


cannot provide immediate isolation in a LOCA. Based upon the results of this evaluation, we have concluded that through the   j use of $~ppropriate manual action, containment isolation can be accomplishbd in a timely manner.
cannot provide immediate isolation in a LOCA.
l 1
Based upon the results of this evaluation, we have concluded that through the j
I (30)   For the other three events, (1) loss of AC power (FSAR $ 15A.1.19), (2) pipe break outside containment (PBOC)
use of $~ppropriate manual action, containment isolation can be accomplishbd in a timely manner.
(steam line break accident) (FSAR $ 15.1.35) and (3) feedwater system piping break (FSAR $ 15.1.37), the reactor would auto-matically isolate. This isolation is not dependent upon the availability of AC power. For all three events, both HPCI and RCIC would be available to provide reactor coolant makeup.
1 (30)
Given the heat capacity of passive heat sinks such as structur-al steel, suppression pool cooling would not be required for about 30 days. Therefore, there is ample time for AC power to be restored. Furthermore, assuming loss of offsite power in the context of pipe breaks outside containment (main steam line break accident and feedwater system break accident) is a con-servatism which stems from the PBOC analysis methodology. That methodology requires the assumption of a loss of offsite power for pipe breaks which result directly in a plant trip of the turbine generator system or reactor protection system. Not-withstanding grid stability analyses, it is assumed that plant trips could cause perturbations of the grid, resulting in the
For the other three events, (1) loss of AC power (FSAR $ 15A.1.19), (2) pipe break outside containment (PBOC)
(steam line break accident) (FSAR $ 15.1.35) and (3) feedwater system piping break (FSAR $ 15.1.37), the reactor would auto-matically isolate.
This isolation is not dependent upon the availability of AC power.
For all three events, both HPCI and RCIC would be available to provide reactor coolant makeup.
Given the heat capacity of passive heat sinks such as structur-al steel, suppression pool cooling would not be required for about 30 days.
Therefore, there is ample time for AC power to be restored.
Furthermore, assuming loss of offsite power in the context of pipe breaks outside containment (main steam line break accident and feedwater system break accident) is a con-servatism which stems from the PBOC analysis methodology.
That methodology requires the assumption of a loss of offsite power for pipe breaks which result directly in a plant trip of the turbine generator system or reactor protection system.
Not-withstanding grid stability analyses, it is assumed that plant trips could cause perturbations of the grid, resulting in the


                                                                          )
)
loss of offsite power. For operation at 5% power or less, how-ever, the turbine generator is not connected to the grid, and thereforb any assumption of induced perturbation to the offsite grid is not valid.
loss of offsite power.
(31) Based on our review of Chapter 15, operation of the plant during low power testing up to levels of 5% of rated power poses no undue risk to the public health and safety. In fact, any risk is substantially less than that already found to be acceptable by the NRC Staff in its review of Chapter 15.
For operation at 5% power or less, how-ever, the turbine generator is not connected to the grid, and thereforb any assumption of induced perturbation to the offsite grid is not valid.
(31)
Based on our review of Chapter 15, operation of the plant during low power testing up to levels of 5% of rated power poses no undue risk to the public health and safety.
In fact, any risk is substantially less than that already found to be acceptable by the NRC Staff in its review of Chapter 15.
Even if the Shoreham TDI diesels are assumed to be unavailable, there is ample assurance that fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded as a result of anticipated operational occurrences, and that the core will be cooled and containment integrity and other vital functions will be maintained in the event of any postulated accident.
Even if the Shoreham TDI diesels are assumed to be unavailable, there is ample assurance that fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded as a result of anticipated operational occurrences, and that the core will be cooled and containment integrity and other vital functions will be maintained in the event of any postulated accident.
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l STATE OF Maryland _)
STATE OF Maryland _)
                                    )   To-wit:
)
To-wit:
COUNTY OF Montgomery)
COUNTY OF Montgomery)
Subscribed to before me this 29tilay       t  of March, 1984.
Subscribed to before me this 29tilay of March, 1984.
as to Glenn G. Sherwood and Eugene C. Tckert ns) OD.             Dr
t as to Glenn G. Sherwood and Eugene C. Tckert ns) OD.
                                                      / Ka ren M. ThompsonNotary Publicf My commission expires:                       July 1, 1986 1
Dr
/ ren M. ThompsonNotary Publicf Ka My commission expires:
July 1, 1986 1
i i
i i
l i
l i


Exhibit 1 FUEL LOAD AND PRECRITICALITY TESTING Chapter 15 Event                   Event Category
Exhibit 1 FUEL LOAD AND PRECRITICALITY TESTING Chapter 15 Event Event Category 1.
: 1. Generator Load Rejection
Generator Load Rejection 2.
* Turbine Trip                                   *
Turbine Trip 3.
:          2.
Turbine Trip with Failure of Generator Breakers to Open 4.
: 3. Turbine Trip with Failure of Generator Breakers to Open                     *
MSIV Closure 5.
: 4. MSIV Closure                                   *
Pressure Regulator Failure - Open 6.
: 5. Pressure Regulator Failure - Open               *
Pressure Regulator Failure - Closed 7.
: 6. Pressure Regulator Failure - Closed             *
Feedwater Controller Failure -
: 7. Feedwater Controller Failure -
Maximum Demand 8.
Maximum Demand                                 ***
Loss of Feedwater Heating 9.
: 8. Loss of Feedwater Heating                       *
Shutdown Cooling (RHR) Malfunction -
: 9. Shutdown Cooling (RHR) Malfunction -
Decreasing Temperature 10.
Decreasing Temperature                         ***
Inadvertent HPCI Pump Start 11.
: 10. Inadvertent HPCI Pump Start                     *
Continuous Control Rod Withdrawal During Power Range Operation Event not possible.
: 11. Continuous Control Rod Withdrawal During Power Range Operation                   *
Component operation possible but Chapter 15 phenomena cannot occur.
* Event not possible.
Event possible but no consequences.
              **  Component operation possible but Chapter 15 phenomena cannot occur.
            ***  Event possible but no consequences.
l
: 12. Continuous Rod Withdrawal During Reactor Startup                      ***
: 13. Control Rod Removal Error During Refueling                            ***
: 14. Fuel' Assembly Insertion Error During Refueling                    ***
: 15. Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location                            *
: 16. Inadvertent Loading and Operation of a Fuel Assembly in Improper Location                            *
: 17. Inadvertent Opening of a Safety / Relief Valve                *
: 18. Loss of Feedwater Flow              ***
: 19. Loss of AC Power                    ***
: 20. Recirculatien Pump Trip              **
: 21. Loss of Condenser Vacuum            *
: 22. Recirculation Pump Seizure          **
: 23. Recirculation Flow Control Failure -
Decreasing Flow                      **
: 24. Recirculation Flow Control Failure With Increasing Flow                **
: 25. Abnormal Startup of Idle Recirculation Pump                  **
l
: 26. Core Coolant Temperature Increcse    ***
: 27. Anticipated Transients Without SCRAM (ATWS)                        *
: 28. Cask Drop Accident
* l      29. Miscellaneous Small Releases Outside Primary Containment          ***
l t
: 30. Off Design Operational Transient as a Consequence of Instrument Line Failure                            ***
: 31. Maln Condenser Gas Treatment
* System Failure                          *
: 32. Liquid Radwaste Tank Rupture            ***
: 33. Control Rod Drop Accident              ***
: 34. Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)  ***
: 35. Pipe Breaks Outside Primary Containment (Steam Line Break Accident)
: 36. Fuel Handling Accident                  ***
: 37. Feedwater System Piping Break          ***
: 38. Failure of Air Ejector Lines
* i l
l
l


Exhibit 2 COLD CRITICALITY TESTING Assumes Un-Event availability Chapter 15 Event                 Category of Offsite AC l
12.
: 1. Generator Load Rejection
Continuous Rod Withdrawal During Reactor Startup 13.
* N/A Turbine Trip                                 *
Control Rod Removal Error During Refueling 14.
: 2.                                                        N/A
Fuel' Assembly Insertion Error During Refueling 15.
: 3. Turbine Trip with Failure of Generator Breakers to Open
Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location 16.
* N/A MSIV Closures                                 *
Inadvertent Loading and Operation of a Fuel Assembly in Improper Location 17.
: 4.                                                        N/A Pressure Regulator Failure - Open             *
Inadvertent Opening of a Safety / Relief Valve 18.
: 5.                                                        N/A
Loss of Feedwater Flow 19.
: 6. Pressure Regulator Failure - Closed
Loss of AC Power 20.
* N/A
Recirculatien Pump Trip 21.
: 7. Feedwater Controller Failure -
Loss of Condenser Vacuum 22.
Maximum Demand                               **      No Loss of Feedwater Heating                     *
Recirculation Pump Seizure 23.
: 8.                                                        N/A
Recirculation Flow Control Failure -
: 9. Shutdown Cooling (RHR) Malfunction -
Decreasing Flow 24.
Decreasing Temperature                       **      No
Recirculation Flow Control Failure With Increasing Flow 25.
: 10. Inadvertent HPCI Pump Start
Abnormal Startup of Idle Recirculation Pump l
* N/A
26.
: 11. Continuous Control Rod Withdrawal During Power Range Operation
Core Coolant Temperature Increcse 27.
* N/A
Anticipated Transients Without SCRAM (ATWS) 28.
* Event not possible.
Cask Drop Accident l
        **  Event possible but essentially no consequences.
29.
Miscellaneous Small Releases Outside Primary Containment t
 
30.
Off Design Operational Transient as a Consequence of Instrument Line Failure 31.
Maln Condenser Gas Treatment System Failure 32.
Liquid Radwaste Tank Rupture 33.
Control Rod Drop Accident 34.
Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident) 35.
Pipe Breaks Outside Primary Containment (Steam Line Break Accident) 36.
Fuel Handling Accident 37.
Feedwater System Piping Break 38.
Failure of Air Ejector Lines i
l l
 
Exhibit 2 COLD CRITICALITY TESTING Assumes Un-Event availability Chapter 15 Event Category of Offsite AC l
1.
Generator Load Rejection N/A 2.
Turbine Trip N/A 3.
Turbine Trip with Failure of Generator Breakers to Open N/A 4.
MSIV Closures N/A 5.
Pressure Regulator Failure - Open N/A 6.
Pressure Regulator Failure - Closed N/A 7.
Feedwater Controller Failure -
Maximum Demand No 8.
Loss of Feedwater Heating N/A 9.
Shutdown Cooling (RHR) Malfunction -
Decreasing Temperature No 10.
Inadvertent HPCI Pump Start N/A 11.
Continuous Control Rod Withdrawal During Power Range Operation N/A Event not possible.
Event possible but essentially no consequences.
i 4
i 4
: 12. Continuous Rod Withdrawal During Reactor Startup                       ** No
 
: 13. Control Rod Removal Error During Re' fueling                         **  No
12.
: 14. Fuel' Assembly Insertion Error During Refueling                     **  No
Continuous Rod Withdrawal During Reactor Startup No 13.
: 15. Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location                             **  No
Control Rod Removal Error During Re' fueling No 14.
: 16. Inadvertent Loading and Operation           l of a Fuel Assembly in Improper               ;
Fuel' Assembly Insertion Error During Refueling No 15.
Location                             **  No ;
Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location No 16.
: 17. Inadvertent Opening of a Safety / Relief Valve
Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.
* N/A
Inadvertent Opening of a Safety / Relief Valve N/A 18.
: 18. Loss of Feedwater Flow               **  No
Loss of Feedwater Flow No 19.
: 19. Loss of AC Power                     **  Yes
Loss of AC Power Yes 20.
: 20. Recirculation Pump Trip             **  No
Recirculation Pump Trip No 21.
: 21. Loss of Condenser Vacuum
Loss of Condenser Vacuum N/A 22.
* N/A
Recirculation Pump Seizure No 23.
: 22. Recirculation Pump Seizure           **  No
Recirculation Flow Control Failure -
: 23. Recirculation Flow Control Failure -
Decreasing Flow No 24.
Decreasing Flow                     **  No
Recirculation Flow Control Failure With Increasing Flow No 25.
: 24. Recirculation Flow Control Failure With Increasing Flow                 **  No
Abnormal Startup of Idle Recirculation Pump No 26.
: 25. Abnormal Startup of Idle Recirculation Pump                   **  No
Core Coolant Temperature Increase No 27.
: 26. Core Coolant Temperature Increase   **  No
Anticipated Transients Without SCRAM (ATWS)
: 27. Anticipated Transients Without SCRAM (ATWS)                         **  No
No 28.
: 28. Cask Drop Accident
Cask Drop Accident N/A 29.
* N/A
Miscellaneous Small Releases Outside Primary Containment No l
: 29. Miscellaneous Small Releases Outside Primary Containment         **  No l
l
l
: 30. Off Design Operational Transient as a Consequence of Instrument Line Failure                              ** No
: 31. Mk'inCondenserGasTreatment                        '
System Failure                                N/A
: 32. Liquid Radwate Tank Rupture                ** No
: 33. Control Rod Drop Accident                  ** No
: 34. Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)    ** Yes
: 35. Pipe Breaks Outside Primary Containment (steam line break accident)
* N/A
: 36. Fuel Handling Accident                    ** No
: 37. Feedwater System Piping Break              ** Yes
: 38. Failure of Air Ejector Lines
* N/A


Exhibit 3
30.
                        .              5% POWER Assumes Un-Event      availability Chapter 15 Event                Category    of Offsite AC 1
Off Design Operational Transient as a Consequence of Instrument Line Failure No 31.
:          1. Generator Load Rejection
Mk'inCondenserGasTreatment System Failure N/A 32.
* N/A
Liquid Radwate Tank Rupture No 33.
: 2. Turbine Trip                              **            No
Control Rod Drop Accident No 34.
: 3. Turbine Trip with Failure of Generator Breakers to Open
Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)
* N/A
Yes 35.
: 4. MSIV Closures                            **              No
Pipe Breaks Outside Primary Containment (steam line break accident)
: 5. Pressure Regulator Failure - Open          **            No
* N/A 36.
: 6. Pressure Regulator Failure - Closed      **              No
Fuel Handling Accident No 37.
: 7. Feedwater Controller Failure -
Feedwater System Piping Break Yes 38.
Maximum Demand                            **              No
Failure of Air Ejector Lines N/A
: 8. Loss of Feedwater Heating                **              No
: 9. Shutdown Cooling (RHR) Malfunction -
Decreasing Temperature                    **              No
: 10. Inadvertent HPCI Pump Start              **              No
: 11. Continuous Control Rod Withdrawal During Power Range Operation              **              No
* Event cannot occur.
          **    Bounded by same event at higher power level per FSAR Chapter 15.


i I
Exhibit 3 5% POWER Assumes Un-Event availability Chapter 15 Event Category of Offsite AC 1
: 12. Continuous Rod Withdrawal During Reactor Startup                     **  No
1.
: 13. Control Rod Removal Error During Refueling                           **  No
Generator Load Rejection N/A 2.
: 14. Fuel' Assembly Insertion Error During Refueling                     **  No
Turbine Trip No 3.
: 15. Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly Into an Improper Location                             **  No
Turbine Trip with Failure of Generator Breakers to Open N/A 4.
: 16. Inadvertent Loading and Operation of a Fuel Assembly in Improper Location                             **  No
MSIV Closures No 5.
: 17. Inadvertent Opening of a Safety / Relief Valve               **  No
Pressure Regulator Failure - Open No 6.
: 18. Loss of Feedwater Flow               **  No
Pressure Regulator Failure - Closed No 7.
: 19. Loss of AC Power                     **  Yes
Feedwater Controller Failure -
: 20. Recirculation Pump Trip             **  No
Maximum Demand No 8.
: 21. Loss of Condenser Vacuum             **  No
Loss of Feedwater Heating No 9.
: 22. Recirculation Pump Seizure           **  No
Shutdown Cooling (RHR) Malfunction -
: 23. Recirculation Flow Control Failure -
Decreasing Temperature No 10.
Decreasing Flow               .
Inadvertent HPCI Pump Start No 11.
                                                  **  No
Continuous Control Rod Withdrawal During Power Range Operation No Event cannot occur.
: 24. Recirculation Flow Control Failure -
Bounded by same event at higher power level per FSAR Chapter 15.
l With Increasing Flow                 **  No
 
: 25. Abnormal Startup of Idle Recirculation Pump                   **  No
i.
: 26. Core Coolant Temperature Increase   **  No
I 12.
: 27. Anticipated Transients Without SCRAM (ATWS)                         **  No
Continuous Rod Withdrawal During Reactor Startup No 13.
: 28. Cask Drop Accident
Control Rod Removal Error During Refueling No 14.
* N/A t
Fuel' Assembly Insertion Error During Refueling No 15.
: 29. Miscellaneous Small Releases Outside Primary Containment         **  No t
Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly Into an Improper Location No 16.
Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.
Inadvertent Opening of a Safety / Relief Valve No 18.
Loss of Feedwater Flow No 19.
Loss of AC Power Yes 20.
Recirculation Pump Trip No 21.
Loss of Condenser Vacuum No 22.
Recirculation Pump Seizure No 23.
Recirculation Flow Control Failure -
Decreasing Flow No 24.
Recirculation Flow Control Failure -
l With Increasing Flow No 25.
Abnormal Startup of Idle Recirculation Pump No 26.
Core Coolant Temperature Increase No 27.
Anticipated Transients Without SCRAM (ATWS)
No 28.
Cask Drop Accident N/A t
29.
Miscellaneous Small Releases Outside Primary Containment No t
l l
l l
l
l
: 30. Off Design Operational Transient as a Consequence of Instrument Line Failure                                                                        ** No
: 31. Md'in Condenser Gas Treatment System Failure                                                                      ** No
: 32. Liquid Radwaste Tank Rupture                                                        ** No
: 33. Control Rod Drop Accident                                                          ** No
: 34. Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)                                                Yes
: 35. Pipe Breaks Outside Primary Containment (Steam Line Break Accident) **                                            Yes
: 36. Fuel Handling Accident                                                              ** No
: 37. Feedwater System Piping Break                                                      ** Yes
: 38. Failure of Air Ejector Lines                                                        ** No


o ECCS LOCA EVALUATIONS 10 CFR E 50.46 Limits Co re         Peak Rod     Time to 10             PCT             Local           Core Wide     ..
30.
Avg. Powe r       MAPLHCR       CFR i 50.46         (F')               Oxidation         Oxidation       *
Off Design Operational Transient as a Consequence of Instrument Line Failure No 31.
(1 or rated)     (kW/Ft1     Limits (mini     flimit 2200')         (Limit 171)       (Limit 111 5.0               1.34             55             2200                 6.5           less than 0.9 2.5               0.67           124             2200                 8.4           less than 1.0 1.25               0.34           285             2100                 9.0           1.0
Md'in Condenser Gas Treatment System Failure No 32.
  .5                 0.13
Liquid Radwaste Tank Rupture No 33.
* 700             2000                 9.0           1.0 ASSUMPTIONS: 10 CFR 50 Appendix K (Standard FSAR Basis)
Control Rod Drop Accident No 34.
Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)
Yes 35.
Pipe Breaks Outside Primary Containment (Steam Line Break Accident) **
Yes 36.
Fuel Handling Accident No 37.
Feedwater System Piping Break Yes 38.
Failure of Air Ejector Lines No
 
o ECCS LOCA EVALUATIONS 10 CFR E 50.46 Limits Co re Peak Rod Time to 10 PCT Local Core Wide Avg. Powe r MAPLHCR CFR i 50.46 (F')
Oxidation Oxidation (1 or rated)
(kW/Ft1 Limits (mini flimit 2200')
(Limit 171)
(Limit 111 5.0 1.34 55 2200 6.5 less than 0.9 2.5 0.67 124 2200 8.4 less than 1.0 1.25 0.34 285 2100 9.0 1.0
.5 0.13 700 2000 9.0 1.0 ASSUMPTIONS: 10 CFR 50 Appendix K (Standard FSAR Basis)
Initial Conditions Based on Equivalent Core at Designated Core Average Power a
Initial Conditions Based on Equivalent Core at Designated Core Average Power a
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Latest revision as of 13:59, 13 December 2024

Affidavit of GG Sherwood,As Rao & EC Eckert in Support of 840320 Supplemental Motion for Low Power OL
ML20087M981
Person / Time
Site: Shoreham File:Long Island Lighting Company icon.png
Issue date: 03/22/1984
From: Eckert E, Appajosula Rao, Sherwood G
GENERAL ELECTRIC CO.
To:
Shared Package
ML20087M974 List:
References
OL, OL-4, NUDOCS 8404020252
Download: ML20087M981 (68)
v • d • e


Text

r UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board I'

In the Matter of

)

)

LONG ISLAND LIGHTING COMPANY

)

Docket No. 50-322 (OL)

)

(Shoreham Nuclear Power Station,

)

Unit 1)

)

AFFIDAVIT OF DR. GLENN G.

SHERWOOD, DR. ATAMBIR S.

RAO, AND MR. EUGENE C.

ECKERT Glenn G.

Sherwood, Atambir S. Rao, and Eugene C. Eckert being duly sworn, depose and state as follows:

(1)

My name is Glenn G.

Sherwood.

I am employed by the General Electric Company as Manager, Safety and Licensing Operation.

My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.

I have been employed in this position since 1976.

My responsibilities in-clude supervision of the preparation of licensing submittals for General Electric BWRs, including analyses performed in Chapter 15 of safety analysis reports.

In particular, I have been involved in the supervision of licensing matters for the Shoreham Nuclear Power Station since the initial submittal of the Shoreham Final Safety Analysis Report (FSAR).

In this re-gard, I am familiar with the analyses performed in Chapter 15' of that document.

From 1974, when I joined General Electric, 8404020252 840331 PDR ADOCK 05000322 O

PDR a;

,1

> to 1976, I was the Manager, Program Control Section.

My responsibilities included managing engineering and i

manufacturing work flow for General Electric's nuclear group.

I have a Bachelor of Science degree in Engineering from the U.S. Naval Academy and a Ph.D. in Engineering from the Univer-sity of Michigan.

(2)

My name is Atambir S. Rao.

I am employed by the General Electric Company as Manager, Plant Safety Systems Engi-neering.

My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.

I was appointed to my present position in 1984.

My responsibilities include ECCS performance analysis, containment performance response analy-sis, and plant safety performance evaluations, including FSAR safety analyses.

I have previously held a number of positions relating to accident and transient analyses since I first 1

joined General Electric in 1973.

Earlier responsibilities have included modeling and analyzing the thermal hydraulic _ behavior of BWR fuel following loss of coolant accidents, assessing the implication of advances in heat transfer, fluid mechanics, thermodynamics and two-phase flow on overall BWR system re-sponse during transients and loss of coolant accidents, devel-l oping emergency operator guidelines, and assessing containment thermal hydraulic and radiological response for various I

r-o

~ accidents and transients.

I have been assigned as Manager, O

Emergency Core Cooling Systems (ECCS) Engineering (1979-80),

and Manager, Containment and Radiological Engineering (198"-84).- I received a Ph.D and a Masters degree in Mechani-2 cal Engineering from the University of California, Berkeley, and a Bachelor of Technology in Mechanical Engineering from the Indian Institute of Technology, Kanpur, India.

(3)

My name is Eugene C.

Eckert.

I am employed by the General Electric Company as Manager, Power Transient Per-forming Engineering, a position I have held since 1971.

My business address is General Electric Company, 175 Curtner Ave-nue, San Jose, California 95125.

I am responsible for estab-lishing the simulation requirements of the computer models needed to perform transient analyses, development of design procedures evaluation of BWR stability, and evaluation and specification of the functional protection systems required for reactor abnormal transient protection.

Immediately upon joining General Electric Company in September 1959, I partici-pated in assignments which included large jet engine control design, aircraft nuclear propulsion control analysis, nuclear submarine kinetics and control analysis, and industrial control simulation analysis at GE's Research and Development Center.

In 1962, I joined General Electric's Nuclear Energy Division to

> i work on Boiling Water Reactor simulation and dynamic analysis.

I have been responsible for design and licensing documentation of the 5 namic analysis for several GE BWRs and have partici-pated in initial startup testing of many of the units.

I re-ceived a Bachelor of Science Degree in Electrical Engineering from Valparaiso University in Indiana in 1958.

I attended Stanford University under an Oak Ridge Fellowship and received a Master of Science Degree in Engineering Science in August 1959.

(4)

Chapter 15 of the Shoreham FSAR provides the re-sults of analyses for the spectrum of accident and transient events that must be accommodated by the Shoreham plant to dem-onstrate compliance with the NRC's regulations.

This portion of the safety analysis is performed to evaluate the ability of the plant to operate without undue risk to the health and safe-ty of the public.

The Shoreham FSAR was submitted to the NRC Staff for review and has been approved by the Staff in its Safety Evaluation Report for Shoreham (NUREG-0420).

(5)

At the request of the Long Island Lighting Compa-ny, General Electric, in conjunction with cognizant LILCO and Stone & Webster personnel, has reviewed all of the events con-sidered in Chapter 15 of the FSAR to determine the effect on-public health and safety of the operation of the Shoreham plant i

, a during fuel load, criticality testing and low power operations.

Although the FSAR considers all phases of the operation of the plant f$om fuel load to operation at 100% power, this review was performed specifically to confirm that operation of the Shoraham plant during low power operation will pose no undue risk to public health and safety.

The review of Chapter 15 was divided into three parts: (1) fuel load and precriticality testing (Phase I), (2) cold criticality testing (Phase II), and (3) low power testing up to 5% of rated power (Phases III and IV).1/

The review was based upon the same criteria and bases as the original Chapter 15 analyses.

Where assumption of a loss or unavailability of offsite power was required in the original analyses, potential unavailability of the TDI diesel generators was considered in this review.

(6)

The General Electric review of Chapter 15 con-firms that operation during the phases identified above will not result in any undue risk to the public health and safety.

In fact, the risk from any Chapter 15 event during both the fuel load and precriticality phase and the cold criticality l

testing phase is essentially non-existent.

The risk to the 1/

Parts (1) and (2) correspond to Phases I and II, respec-tively, as described in the Affidavit of Messrs. Notaro and Gunther.

Part (3) corresponds to Phases III and IV, combined, as described in that affidavit.

l i

l

- public health and safety from the Chapter 15 avents postulated for low power testing up to 5% of rated power is small in com-parison'to the risks already found acceptable for 100% power operation.' As already indicated, this review considered the impact of potential diesel unavailability.

Phase I:

Fuel Loading and Precriticality Testing (7)

This phase of operation of the Shoreham plant in-cludes only initial fuel loading and precriticality testing.

The reactor will remain at essentially ambient temperature and atmospheric pressure.

The reactor will not be taken critical.

Any increase in temperature beyond ambient conditions will be due only to external heat sources such'as recirculation pump heat.

There will be no heat generation in the core.

Details of the steps to be performed during these operations are de-scribed in the Phase I discussion in the affide At submitted by Messrs. Notaro and Gunther.

(8)

The review of the Chapter 15 analysis revealed that of the 38 accident or transient events addressed in Chap-ter 15, 18 of the events could not occur during Phase I because of the operating conditions of the plant.

An additional 5 events could physically occur, but given the plant conditions, could rat constitute events in the context of the Chapter 15

l a

J i

safety analysis.

The remaining 15 events could possibly occur, although occurrence is highly unlikely given the plant condi-

^

tions.

In any event, it is readily apparent that the potential consequences of these 15 events would be trivial.

Exhibit 1 below lists the category into which each Chapter 15 event j

falls.

(9)

The 18 Chapter 15 events which could not occur i

j during Phase I are precluded by the operating conditions of the 1

reactor.

These events all involve operating modes or component 1

operation which are not possible during this phase.

For exam-i l

ple, during fuel loading and precriticality testing, the reac-1 tor is at essentially ambient temperature and atmospheric pres-i sure.

Accordingly, no steam is available.

Thus, all events which would require pressurized conditions are precluded.

Events such as turbine trip (FSAR 5 15A.1.2), loss of feedwater heating (FSAR 5 15A.1.8) and inadvertent opening of a safety j

relief valve require the generation of steam for the event to occur.

Similarly, there is no steam flow to interrupt, thus precluding an MSIV closure event (FSAR 5 15A.1.4).

Other i

1 events are precluded by definition.

Thus, events such as con-tinuous control rod withdrawal during power range operation (FSAR 5 15A.1.11) and operation of a fuel assembly in an.im-i proper location (FSAR 5 15A.1.16) cannot be. postulated.

l t

. _. - ~,

, i (10)

In addition to the 18 events which nimply cannot occur, there are 5 events for which the component operation evaluated in Chapter 15 could occur, but the phenomena of in-terest in Chapter 15 could not exist.

All recirculation pump events, such as recirculation pump trip (FSAR $ 15A.1.20) and abnormal startup of an idle recirculation pump (FSAR 5 15A.1.25), would be of interest only if they could affect core physics or thermal-hydraulic conditions.

With no heat generation or boiling in the core, there are no pertinent phe-nomena (such as temperature differences or void collapses) to evaluate.

Another example, the core coolant temperature in-crease event (FSAR $ 15.A.1.26), postulates a loss of RHR cool-ing.

Even if the RHR system was operated in Phase I, there would be no temperature increase from decay heat to evaluate should the RHR system be lost.

(11)

The remaining 15 events addressed in Chapter 15 could possibly occur.

However, our review established that all are trivial events which have no potential to impact public health and safety.

Prior to initial criticality, there are no fission products in the core and no decay heat exists.

It fol-lows that core cooling is not required.

In addition, with no fission product inventory, there are no fission product re-leases possible.

Thus, for reactor events such as a control

, ]

rod removal error (FSAR 5 15A.1.13) and a control rod drop (FSAR $ 15.1.33) and for non-reactor events such as a fuel han-dling abcident (FSAR $ 15.1.36) or a liquid radwaste tank rup-ture (FSAR'l 15.1.32), there could be no radiological conse-quences.

Therefore, there is no risk to public health and safety.

(12)

Even a loss of coolant accident (FSAR $ 15.1.34) could have no radiological consequences during Phase I.

No core cooling is required.

No fission product release is possi-ble.

The fuel simply could not be challenged by a complete draindown of the reactor vessel for an unlimited period of time.

(13)

In summary, the review of Chapter 15 events for fuel loading and precriticality testing indicates that nany Chapter 15 events simply cannot occur, and for those that can, there can be no radiological consequences.

Therefore, there is no possible risk to the public health and safety.

This conclu-sion is not affected by any postulated diesel generator unavailability because it is in no way dependent on the avail-ability or unavailability of any AC power.

Phase II:

Cold Criticality Testing (14)

This phase of low power testing of the Shoreham

1 l

4 "

plant will include cold criticality testing of the plant at es-sentially ambient temperature and atmospheric pressure.

The power 1 vel during this phase of testing will be in the range of.0001% to.001% of rated power.

Details of the testing to l

be performed during this phase are described in the Notaro Af-fidavit.

1 (15)

The review of Chapter 15 revealed that of the.38 accident or transient events included there, 15 of the events could not occur because of the operating conditions of the plant during Phase I.

See Exhibit 2.

A number of these events are not possible because the reactor will be at essentially ambient temperature and pressure and no steam will be gener-ated.

For example, the generator load rejection event (FSAR

$ 15A.1.1) could not occur during this testing phase because steam is needed to drive the main turbine generator to permit connecting it to the LILCO transmission system.

Another exam-ple, the loss of condenser vacuum event (FSAR 5 15A.1.21),

could not occur because it assumes that. steam is available to draw a vacuum in the main condenser.

A third example, the in-advertent HPCI pump start event (FSAR 5 15A.1.10), could not occur because there will be no steam available to power the HPCI pump, a steam driven ECCS pump.

Other Chapter 15 events could not occur because.they are precluded by the configuration e

m

_11 of the plant during this phase of low po.wer testing.

An exam-ple of this type of event is the MSIV closure (FSAR 5 15A.1.4).

The MSI s will normally be closed throughout all of the op-erations conducted during this phase of low power testing.

In any event, there is no steam generated by the reactor to flow through the steam lines.

(16)

In addition to the 15 events that could not occur during Phase I, many of the 23 events remaining in the Chapter 15 analysis are far less likely to occur during low power testing than during normal operations.

For example, the recirculation pump trip (FSAR $ 15A.1.20), the recirculation pump seizure (FSAR 5 15A.1.22), the recirculation flow control failures (FSAR 9 15A.1.23 and 24) and the abnormal startup of

)

idle recirculation pump (FSAR $ 15A.1.25) events, although physically possible, are not as likely to occur because the recirculation pumps are used for only limited periods of time during this phase of the testing program.

Similarly, the loss of feedwater event (FSAR $ 15A.1.18) is very unlikely because little, if any, make-up water will have to be supplied to the reactor.

Moreover, make-up water would not normally be supplied by the feedwater system.under these conditions.

Other very unlikely events include miscellaneous.small releases out-side primary containment (FSAR $ 15.1.29), off design l

l

'l

=.

. l operational transient as a consequence of instrument line fail-I ure (FSAR $ 15.1.30), and feedwater system piping break (FSAR

$ 15.1.3 ).

Thus, many of the Chapter 15 events that are phys-ically possible during Phase II remain very unlikely in light of the plant conditions that will then exist.

(17)

Nonetheless, all 23 possible events contained in Chapter 15 were reviewed to reaffirm that the consequences of these events, should one occur during Phase I of low power testing, would be bounded by the consequences analyzed for the event considered in the FSAR.

A discussion of some of the 23 possible events contained in Chapter 15 illustrates the basis for this conclusion.

The continuous control rod withdrawal during startup event (FSAR $ 15A.1.12) is applicable to op-eration in the power, source and/or intermediate range of op-eration.

During cold functional criticality testing, the reac-tor will operate in the source and intermediate ranges and therefore the conclusions contained in Chapter 15 are applica-ble to this event should it occur during this phase of low power testing.

As the FSAR indicates, this event would not re-sult in any rele.e ' af radioactive material from the fuel at any power level.

Another example is the fuel handling accident (FSAR $ 15.1.36).

As stated in the FSAR, the most severe fuel handling accident from a radiological viewpoint is a dropping o

r r

v--

7

-__..,e

_ = -

- of the fuel assembly onto the top of the core.

The FSAR analy-l sis assumes that the fuel contains a fission product inventory i

l equivalent to operation of 1000 days at full rated power.

This assumption results in an equilibrium fission product concentra-I tion at the time the reactor is shut down.

But as already noted, the fission product inventories in the core will be sig-nificantly less during Phase II low power testing than the in-ventories analyzed in the FSAR because of the extremely low power levels (.0001% to.001% of rated power) achieved during this testing.

Thus, even if a handling accident took place and fuel damage did occur, there would be significantly less fis-sion products to be released from the core than those that have

}

already been analyzed and found acceptable in the FSAR.

A third example is the liquid radwaste tank rupture event (FSAR $

15.1.32).

This event assumes the rupture of a radwaste tank that contains a substantial amount of contaminated liquids gen-erated during the operation of the reactor.

But again, since Phase II low power testing results in insignificant power lev-els in the reactor, there will be little, if any, radioactive i

liquids in the radwaste tank should such a rupture occur.

I

]

Thus, even the minimal consequences already described in the

'FSAR for the design basis event would be further reduced under these low power testing conditions.

For each of these events, the review concluded that the consequences are significantly 1

,.e T

T

, less severe for any event occurring during the cold functional criticality testing than for the event analyzed in Chapter 15.

To summa ize, because of the extremely low power levels reached during this' testing phase, fission product inventory in the core will be only a small fraction of that assumed for the Chapter 15 analyses.

As indicated above, the FSAR assumes op-eration at 100% power for 1000 days in calculating fission product inventory; the inventory during Phase II low power testing will be less than one one-hundred-thousandth (.00001) of the fission product inventory assumed in the FSAR.

Conse-quently, none of the events analyzed in Chapter 15 could result in a release of radioactivity during cold criticality testing that would harm the public health and safety.

(18)

The review of Chapter 15 events for Phase II testing and the conclusions reached are unaffected by any unavailability of the TDI diesels.

Of the 23 possible Chapter 15 events reviewed, 20 of the events in the FSAR do not require the assumption of loss or unavailability of offsite AC power.

See Exhibit 2.

Thus, our conclusions for these 20 of the 23 possible events are independent of the status of the diesels.

(19)

The three events that do assume loss or unavailability of offsite AC power are (1) pipe breaks inside the primary containment (LOCA) (FSAR $ 15.1.34), (2) feedwater

__ _ _. ~

4 system piping break (FSAR $ 15.1.37), and (3) the loss of AC power event (FSAR $ 15A.1.19).

With respect to these~ events, I

the LOCI'would be the most limiting event.

The review has shown that'if a LOCA did occur during the cold criticality testing phase, however remote that possibility, there would be time on the order of months available to restore make-up water for core cooling.

At the power levels achieved during Phase l

II, fission product inventory is very low.

At most, decay heat will, on the average, be a fraction of a watt per rod, with no single rod exceeding approximately 2 watts.

This is less, roughly, than the heat output of a Christmas tree bulb.

It 4

follows that the fuel cladding temperature would not exceed the limits of 10 CFR $ 50.46 and Appendix K even after months with-out cooling and without any source of AC power.

(20)

The loss of AC power event (FSAR $ 15A.1.19) and the feedwater system piping break (FSAR 5 15.1.37) under cold criticality testing conditions do not rely on the diesel gener-ators for mitigation of the event.

For these events, since no loss of coolant occurs and the decay heat is minimal, core cooling is achieved. without AC power, using the existing core water inventory and heat losses to ambient, for essentially un-limited periods of time.

In any event, as demonstrated in the Schiffmacher Affidavit, AC power sources can and will be

. readily supplied to the Shoreham plant even if one assumes the simultaneous loss of all three emergency diesel generators.

(21)

In addition to our conclusions that the limiting LOCA event could not approach the limits of 10 CFR $ 50.46 and Appendix K during Phase II low power testing, there are other reasons why our findings with respect to the three events that assume loss of AC power are independent of the availability of the TDI diesels.

The LOCA (pipe break inside containment) and the feedwater system piping break postulate the double ended rupture of a piping system.

Because the reactor will be at es-sentially ambient temperature and atmospheric pressure during Phase II, it is extremely unlikely that such a pipe break would ever occur.

In fact, the NRC Staff does not require double ended ruptures to be postulated for low temperature and low pressure systems in safety analyses.

Thus, these events are much less likely during cold criticality testing than during normal operation.

(22)

The review of Chapter 15 events for cold criti-cality testing indicates that performance of these activities at Shoreham involves essentially no risk to the public health and safety.

This conclusion is not affected by any postulated diesel unavailability.

In fact, even if AC power were not i

available for extended periods of time, fuel design limits and

. design conditions of the reactor coolant pressure boundary would not be approached or exceeded as a result of anticipated operati nal occurrences, and the core would be adequately cooled in the unlikely event of a postulated accident.

Phases III and IV:

Low Power Testing Up To 5% of Rated Power (23)

These aspects of low power testing will include operation of the plant at power levels up to 5% of rated power.

Details of the testing to be performed during this phase of op-eration are described as Phases III and IV in the Notaro Affi-davit.

(24)

The review of the 38 Chapter 15 events for these phases of low power testing operations revealed that two of the events in Chapter 15, generator load rejection (FSAR 5 15A.1.1) and turbine trip with generator breaker failure (FSAR 5 15.1.2) cannot occur because the generator will not be connected to the grid during these phases of testing.

A third event, the cask drop, is precluded by design as stated in FSAR 5 15.1.28.

See Exhibit 3.

(25)

Of the remaining 35 events that can occur during this phase of operation, 31 of the events do not assume loss or unavailability of AC power.

For each of these 31 events,

. operation of the plant up to 5% of rated power will be bounded by the Chapter 15 analysis.

Since the Chapter 15 analysis con-i e,

siders all possible phases of plant operation, it follows that operation a't 5% can result in consequences ~1ess severe than those analyzed in Chapter 15.

For example, the turbine trip event (FSAR $ 15A.1.2) assumes that the limiting event occurs with the reactor operating at 105% of rated steam flow coupled with failure of the turbine bypass valves to open.

Even this limiting event does not result in any fuel failures.

ESAR 5 15A.1.2 specifically notes that turbine trips at power levels less than 30% of rated power are bounded by the limiting analy-sis.

Another example is the loss of feedwater heating event (FSAR $ 15A.1.8).

This event assumes continuous operation of the feedwater system and the most severe possible loss of feedwater heating, resulting in the injection of colder feedwater.

For operation at power levels less than 5%, the im-pact of lost feedwater heating is minimal because of the low feedwater flow.

Since these analyses are not required to as-sume the absence of AC power, potential unavailability of the TDI diesels has no effect on the assessment of these events.

l (26)

Not only are the results of these 31 events bounded by the Chapter 15 analysis, the consequences of these events are also less than the consequences stated in the FSAR.

l

1

_19_

i 1

First, the power limitations during low power testing up to 5%

power, the fission product inventory in the core will not ex-4 b

ceed 5% of the values assumed in the FSAR.

In fact, because of the intermittant type of operations conducted during low power f

testing, equilibrium fission product inventory for even 5%

i power is unlikely to be achieved.

This low fission product in-ventory reduces risk in two ways:

(a) the amount of decay heat present in the core following shutdown is substantially re-duced, and (b) the amount of radioactivity that could be re-leased upon fuel failure is substantially reduced.

(27)

The second factor contributing to the signifi-cantly lower risk during low power operation is the increased I

time available for preventive or mitigating action should such l

action be deemed desirable by the operator.

Longer time is available because the limited power levels mean that it takes longer for the plant to reach setpoints and limits.

For exam-ple, on loss of feedwater (FSAR 5 15A.1.18), the water level in the reactor will decrease at a slower rate than if the event occured at 100% power.

This gives the operator more time to act manually to restore feedwater before an automatic action takes place.

Similarly, in the loss of condenser vacuum event (FSAR $ 15.A.1.21), the operator will have more time to identi-fy the decreasing vacuum and to take steps to remedy the i

i l

. situation before automatic actions such as turbine trip, feedpump trip or main steam isolation occur.

Another example is the in steam isolation valve closure event (FSAR 5 15A.1.4).'

At five percent power, the amount of heat produced upon isolation of the reactor vessel (which is followed by a reactor scram) results in a much slower pressure and tempera-ture increase than would be experienced at 100% power.

This gives the operator more time to manually initiate reactor cool-ing rather than relying on automatic action.

In effect, the operator may end the transient before there is any substantial impact on the plant.

(28)

The third factor contributing to the signifi-cantly lower risk during low power testing is the reduction in the required capacity for mitigating systems.

Because of the lower levels of decay heat present following operation at 5%

power, the demand for core cooling and auxiliary systems is substantially reduced, permitting the operation of fewer sys-tems a'nd components to mitigate any event.

It follows that the AC power requirements for event mitigation are substantially reduced for 5% power operation as compared to 100% power op-eration.

(29)

As already noted, only four of the events ana-lyzed in Chapter 15 require the assumption of the

. unavailability of offsite AC power for operation during Phases I

j III and IV.

Of these four events, the loss of coolant accident e

is the $ost limiting event.

The Chapter 15 LOCA analysis as-sumes the unavailability of offsite AC power.

This is a con-servative licensing assumption.

In fact, as described in de-tail in the Schiffmacher Affidavit, there are multiple sources I

of AC power available to the Shoreham site (e.g.,

emergency diesel generators, two normal sources of offsite power, blackstart gas turbines at Holtsville, Southhold, and East Hampton, a blackstart gas turbine on the Shoreham site, and mo-bile diesel generators).

Thus, AC power will be available at j

Shoreham to mitigate a loss of coolant accident during low power operations up to 5% rated power.

In the unlikely event offsite AC power is lost, it can be restored within sufficient time to prevent exceeding the limits of 10 CFR 5 50.46 and Ap-pendix K.

GE has determined that for 5% power so long as reflooding of the core has occurred within approximately one hour, 5 50.46 criteria will be met.2/

As the Schiffmacher Af-fidavit demonstrates, power can be restored to Shoreham within minutes.

An evaluation has been performed to assure the ade-quacy of containment isolation in the event AC power sources 2/

As shown in the Exhibit 4 below, lower power levels will result in more time to restore power and core cooling for a postulated LOCA.

Thus, for 1% power approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> are available.

i

-_ -.. _ - = _ _ _ _. - -

l

\\

- l d

cannot provide immediate isolation in a LOCA.

Based upon the results of this evaluation, we have concluded that through the use of appropriate manual action, containment isolation can be accomplished in a timely manner.

I (30)

For the other three events, (1) loss of AC power l

l (FSAR $ 15A.1.19), (2) pipe break outside containment (PBOC)

J (steam line break accident) (FSAR 6 15.1.35) and (3) feedwater system piping break (FSAR $ 15.1.37), the reactor would auto-matically isolate.

This isolation is not dependent upon the availability of AC power.

For all three events, both HPCI and RCIC would be available to provide reactor coolant makeup.

Given the heat capacity of passive heat sinks such as structur-4 al steel, suppression pool cooling would not be required for about 30 days.

Therefore, there is ample time for AC power to i

be restored.

Furthermore, assuming loss of offsite power in the context of pipe breaks outside containment (main steam line break accident and feedwater system break accident) is a con-servatism which stems from the PBOC analysis methodology.

That methodology requires the assumption of a loss of offsite power for pipe breaks which result directly in a plant trip of the turbine generator system or reactor protection system.

Not-withstanding grid stability analyses, it is assumed that plant trips could cause perturbations of the grid, resulting in the e

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loss of offsite power.

For operation at 5% power or less, how-ever, the turbine generator is not connected to the grid, and therefo e any assumption of induced perturbation to the offsite grid is not valid.

(31)

Based on our review of Chapter 15, operation of the plant during low power testing up to levels of 5% of rated power poses no undue risk to the public health and safety.

In fact, any risk is substantially less than that already found to be acceptable by the NRC Staff in its review of Chapter 15.

Even if the Shoreham TDI diesels are assumed to be unavailable, there is ample assurance that fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded as a result of anticipated operational occurrences, and that the core will be cooled and containment integrity and other vital functions will be maintained in the event of any postulated accident.

Glenn G. Sherwood M

Gd Atambir S.

Rao Eugene C.

Eckert

. _ - =. _ _.

i i i

MEtuNyc)

STATE OF

)

To-wit:

COUNTY GE lYES7 tee 5722)

Subscribed to before me this h ay of March, 1984.

n"u eM Notary p lic JOSE M. TEJAD1 Notary Pubhc. State et r;e., y.n My commission expires:

Ns. 03 4727173 yualihed in Srcas County Cert. Filed in Westchester (a -

Commission Empires March 30,19[fy.

i 1

.1 i

a n

~

Exhibit 1

, FUEL LOAD AND PRECRITICALITY TESTING Chapter 15 Event Event Category 1.

Generator Load Rejection 2.

Turbine Trip 3.

Turbine Trip with Failure of Generator Breakers to Open 4.

MSIV Closure 5.

Pressure Regulator Failure - Open 6.

Pressure Regulator Failure - Closed 7.

Feedwater Controller Failure -

Maximum Demand 8.

Loss of Feedwater Heating 9.

Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature 10.

Inadvertent HPCI Pump Start 11.

Continuous Control Rod Withdrawal During Power Range Operation Event not possible.

Component operation possible but Chapter 15 phenomena cannot occur.

Event possible but no consequences.

-l l

~

. 12.

Continuous Rod Withdrawal During Reactor Startup 13.

Control Rod Removal Error During Rdfueling 14.

Fuel-Assembly Insertion Error During Refueling 15.

Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location 16.

Inadvertent Loading and Operation of a Fuel Assembly in Improper Location 17.

Inadvertent Opening of a Safety / Relief Valve 18.

Loss of Feedwater Flow 19.

Loss of AC Power 20.

Recirculation Pump Trip 21.

Loss of Condenser Vacuum 22.

Recirculation Pump Seizure 23.

Recirculation Flow Control Failure -

Decreasing Flow 24.

Recirculation Flow Control Failure With Increasing Flow 25.

Abnormal Startup of Idle Recirculation Pump 26.

Core Coolant Temperature Increase 27.

Anticipated Transients Without SCRAM (ATWS) 28.

Cask Drop Accident 29.

Miscellaneous Small Releases Outside Primary Containment l

3-30.

Off Design Operational Transient as a Consequence of Instrument Line Failure 31.

Main Condenser Gas Treatment Syste,m Failure 32.

Liquid Radwaste Tank Rupture 33.

Control Rod Drop Accident 34.

Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident) 35.

Pipe Breaks Outside Primary Containment (Steam Line Break Accident) 36.

Fuel Handling Accident 37.

Feedwater System Piping Break 38.

Failure of Air Ejector Lines

_____.____-_._._._______m

_-_____-___-_____-_______-_____m.._

_____m____._-__m______

Exhibit 2 COLD CRITICALITY TESTING Assumes Un-Event availability Chapter 15 Event Category of Offsite AC N/A 1.

Generator Load Rejection N/A 2.

Turbine Trip 3.

Turbine Trip with Failure of Generator Breakers to Open N/A 4.

MSIV Closures N/A 5.

Pressure Regulator Failure - Open N/A 1

6.

Pressure Regulator Failure - Closed N/A

~

7.

Feedwater Controller Failure -

Maximum Demand No 8.

Loss of Feedwater Heating N/A 9.

Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature No 10.

Inadvertent HPCI Pump Start N/A 11.

Continuous Control Rod Withdrawal During Power Range Operation N/A Event not possible.

Event possible but. essentially no consequences.

-e

l i

, 12.

Continuous Rod Withdrawal During No Reactor Startup l

l 13.

Control Rod Removal Error During No Refueling 14.

Fuel Assembly Insertion Error No During Refueling

.15.

Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location No 16.

Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.

Inadvertent Opening of a N/A Safety / Relief Valve 18.

Loss of Feedwater Flow No 19.

Loss of AC Power Yes 20.

Recirculation Pump Trip No a

21.

Loss of Condenser Vacuum N/A 22.

Recirculation Pump Seizure No 23.

Recirculation Flow Control Failure -

Decreasing Flow No 24.

Recirculation Flow Control Failure With Increasing Flow No 1

25.

Abnormal Startup of Idle Recirculation Pump No 26.

Core Coolant Temperature Increase No 27.

Anticipated Transients Without SCRAM (ATWS)

No 28.

Cask Drop Accident N/A 29.

Miscellaneous Small Releases outside Primary Containment No-

. 30.

Off Design Operational Transient as a Consequence of Instrument Line Failure No 31.

Mk'in Condenser Gas Treatment

~

System Failure N/A 32.

Liquid Radwate Tank Rupture No 33.

Control Rod Drop Accident No 34.

Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)

Yes 35.

Pipe Breaks Outside Primary Containment (steam line break accident)

  • N/A 36.

Fuel Handling Accident No 37.

Feedwater System Piping Break Yes 38.

Failure of Air Ejector Lines N/A i

f

.,7

I Exhibit 3 5% POWER Assumes Un-Event availability Chapter 15 Event Category of Offsite AC N/A 1.

Generator Load Rejection 2.

Turbine Trip No 3.

Turbine Trip with Failure of Generator Breakers to Open N/A 4.

MSIV Closures No 5.

Pressure Regulator Failure - Open No 6.

Pressure Regulator Failure - Closed No 7.

Feedwater Controller Failure -

Maximum Demand No 8.

Loss of Feedwater Heating No 9.

Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature No 10.

Inadvertent HPCI Pump Start No 11.

Continuous Control Rod Withdrawal During Power Range Operation No Event cannot occur.

Bounded by same event at higher power level per FSAR Chapter 15.

i i

I

.~

12.

Continuous Rod Withdrawal During Reactor Startup No 13.

Control Rod Removal Error During Ra' fueling No 14.

Fuel-Assembly Insertion Error During Refueling No 15.

Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly Into an Improper Location No 16.

Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.

Inadvertent Opening of a Safety / Relief Valve No 18.

Loss of Feedwater Flow No 19.

Loss of AC Power Yes 20.

Recirculation Pump Trip No 21.

Loss of Condenser Vacuum No 22.

Recirculation Pump Seizure No 23.

Recirculation Flow Control Failure -

Decreasing Flow No 24.

Recirculation Flow Control Failure -

With Increasing Flow No 25.

Abnormal Startup of Idle Recirculation Pump No 26.

Core Coolant Temperature Increase No 27.

Anticipated Transients Without SCRAM (ATWS)

No 28.

Cask Drop Accident

}h0L 29.

Miscellaneous Small Releases Outside Primary Containment No l

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1 30.

Off Design Operational Transient l

as a Consequence of Instrument Line Failure No 31.

Mk'inCondenserGasTreatment

~

No System Failure No 32.

Liquid Radwaste Tank Rupture 33.

Control Rod Drop Accident No 34.

Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)

Yes 35.

Pipe Breaks Outside Primary Containment (Steam Line Break Accident) **

Yes 36.

Fuel Handling Accident No 37.

Feedwater System Piping Break Yes 38.

Failure of Air Ejector Lines No p-

ECCS LOCA EVALUATIONS 10 CFR E 50.46 Limits Core Peak Rod Time to 10 PCT Local Core Wide Avg. Powe r MAPLHCR CFR 5 50.46 (F')

Oxidation Oxidation (1 or rated) ikW/ft1 Limits (min)

(Limit 2200')

flimit 171)

(Limit 111 5.0 1.34 55 2200 6.5 less than 0.9 2.5 0.67 124 2200 8.4 less than 1.0 1.25 0.34 285 2100 9.0 1.0

.5 0.13 700 2000 9.0

1.0 ASSUMPTIONS

10 CFR 50 Appendix K (Standard FSAR Basis)

Initial Conditions Based on Equivalent Core at Designated Core Average Power i

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4 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matt'er of

)

)

LONG ISLAND LIGHTING COMPANY

)

Docket No. 50-322 (OL)

)

(Shoreham Nuclear Power Station,

)

Unit 1)

)

AFFIDAVIT OF DR. GLENN G.

SHERWOOD, DR. ATAMBIR S.

RAO, AND MR. EUGENE C.

ECKERT Glenn G.

Sherwood, Atambir S. Rao, and Eugene C. Eckert being duly sworn, depose and state as follows:

(1)

My name is Glenn G.

Sherwood.

I am employed by the General Electric Company as Manager, Safety and Licensing Operation.

My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.

I have been employed in this position since 1976.

My responsibilities in-clude supervision of the preparation of licensing submittals for General Electric BWRs, including analyses performed in Chapter 15 of safety analysis reports.

In particular, I have been involved in the supervision of licensing matters for the Shoreham Nuclear Power Station since the initial submittal of the Shoreham Final Safety Analysis Report (FSAR).

In this re-gard, I am familiar with the analyses performed in Chapter'15 of that document.

From 1974, when I joined General Electric,

1 s

~2-l to 1976, I was the Manager, Program Control Section.

My T

responsibilities included managing engineering and manufacturing work flow for General Electric's nuclear group.

I have a Bachelor of Science degree in Engineering from the U.S. Naval Academy and a Ph.D. in Engineering from the Univer-sity of Michigan.

(2)

My name is Atambir S. Rao.

I am employed by the General Electric Company as Manager, Plant Safety Systems Engi-neering.

My business address is General Electric Company, 175 Curtner Avenue, San Jose, California 95125.

I was appointed to my present position in 1984.

My responsibilities include ECCS performance analysis, containment performance response analy-sis, and plant safety performance evaluations, including FSAR safety analyses.

I have previously held a number of positions relating to accident and transient analyses since I first joined General Electric in 1973.

Earlier responsibilities have included modeling and analyzing the thermal hydraulic behavior of BWR fuel following loss of coolant accidents, assessing the implication of advances in heat transfer, fluid mechanics, thermodynamics and two-phase flow on overall BWR system re-sponse during transients and loss of coolant accidents, devel-oping emergency operator guidelines, and assessing containment thermal hydraulic and radiological response for various

)

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s accidents and transients.

I have been assigned as Manager, Emergency Core Cooling Systems (ECCS) Engineering (1979-80),

and Mankger, Containment and Radiological Engineering (1982-84).' I received a Ph.D and a Masters degree in Mechani-cal Engineering from the University of California, Berkeley, and a Bachelor of Technology in Mechanical Engineering from the Indian Institute of Technology, Kanpur, India.

(3)

My name is Eugene C.

Eckert.

I am employed by the General Electric Company as Manager, Power Transient Per-forming Engineering, a position I have held since 1971.

My business address is General Electric Company, 175 Curtner Ave-nue, San Jose, California 95125.

I am responsible for estab-lishing the simulation requirements of the computer models needed to perform transient analyses, development of design procedures evaluation of BWR stability, and evaluation and specification of the functional protection systems required for reactor abnormal transient protection.

Immediately upon joining General Electric Company in September 1959, I partici-pated in assignments which included large jet engine control design, aircraft nuclear propulsion control analysis, nuclear submarine kinetics and control analysis, and industrial control simulation analysis at GE's Research and Development Center.

In 1962, I joined General Electric's Nuclear Energy Division to l

s

. 4 work on Boiling Water Reactor simulation and dynamic analysis.

I have been responsible for design and licensing documentation of the dynamic analysis for several GE BWRs and have partici-pated in initial startup testing of many of the units.

I re-ceived a Bachelor of Science Degree in Electrical Engineering from Valparaiso University in Indiana in 1958.

I attended Stanford University under an Oak Ridge Fellowship and received a Master of Science Degree in Engineering Science in August 1959.

(4)

Chapter 15 of the Shoreham FSAR provides the re-sults of analyses for the spectrum of accident and transient events that must be accommodated by the Shoreham plant to dem-onstrate compliance with the NRC's regulations.

This portion of the safety analysis is performed to evaluate the ability of the plant to operate without undue risk to the health and safe-ty of the public.

The Shoreham FSAR was submitted to the NRC Staff for review and has been approved by the Staff in its Safety Evaluation Report for Shoreham (NUREG-0420).

(5)

At the request of the Long Island Lighting Compa-l ny, General Electric, in conjunction with cognizant LILCO and i

Stone & Webster personnel, has reviewed all of the events con-t cidered in Chapter 15 of the FSAR to determine the effect on public health and safety of the operation of the Shoreham plant

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during fuel load, criticality testing and low power operations.

Although the FSAR considers all phases of the operation of the plant from fuel load to operation at 100% power, this review I

was performed specifically to confirm that operation of the Shoreham plant during low power operation will pose no undue risk to public health and safety.

The review of Chapter 15 was divided into three parts: (1) fuel load and precriticality testing (Phase I), (2) cold criticality testing (Phase II), and (3) low power testing up to 5% of rated power (Phases III and IV).1/

The review was based upon the same criteria and bases as the original Chapter 15 analyses.

Where assumption of a loss or unavailability of offsite power was required in the original analyses, potential unavailability of the TDI diesel generators was considered in this review.

(6)

The General Electric review of Chapter 15 con-firms that operation during the phases identified above will l

not result in any undue risk to the public health and safety.

i In fact, the risk from any Chapter 15 event during both the fuel load and precriticality phase and the cold criticality testing phase is essentially non-existent.

The risk to the i

l l

1/

Parts (1) and (2) correspond to Phases I and II, respec-tively, as described in the Affidavit of Messrs. Notaro and Gunther.

Part (3) corresponds to Phases III and IV, combined, as described in that affidavit.

1 e

i,

public health and safety from the Chapter 15 events postulated for low power testing up to 5% of rated power is small in com-I parison to the risks already found acceptable for 100% power operation.' As already indicated, this review considered the impact of potential diesel unavailability.

Phase I:

Fuel Loading and Precriticality Testing (7)

This phase of operation of the Shoreham plant in-cludes only initial fuel loading and precriticality testing.

The reactor will remain at essentially ambient temperature and atmospheric pressure.

The reactor will not be taken critical.

Any increase in temperature beyond ambient conditions will be due only to external heat sources such as recirculation pump heat.

There will be no heat generation in the core.

Details of the steps to be performed during these operations are de-scribed in the Phase I discussion in the affidavit submitted by Messrs. Notaro and Gunther.

(8)

The review of the Chapter 15 analysis revealed that of the 38 accident or transient events addressed in Chap-ter 15, 18 of the events could not occur during Phase I because of the operating conditions of the plant.

An additional 5 events could physically occur, but given the plant conditions, could not constitute events in the context of the Chapter 15 l

s 1

3 i

i safety analysis.

The remaining 15 events could possibly occur, 2

although occurrence is highly unlikely given the plant condi-2 tions. EIn any event, it is readily apparent that the potential i

consequences of these 15 events would be trivial.

Exhibit 1 l

below lists the category into which each Chapter 15 event l

falls.

j (9)

The 18 Chapter 15 events which could not occur l

l during Phase I are precluded by the operating conditions of the 4

reactor.

These events all involve operating modes or component operation which are not possible during this phase.

For exam-t ple, during fuel loading and precriticality testing, the reac-tor is at essentially ambient temperature and atmospheric pres-sure.

Accordingly, no steam is available.

Thus, all events which would require pressurized conditions are precluded.

Events such as turbine trip (FSAR $ 15A.1.2), loss of feedwater l

heating (FSAR S 15A.1.8) and inadvertent opening of a safety relief valve require the generation of steam for the event to l

occur.

Similarly, there is no steam flow to interrupt, thus i

precluding an MSIV closure event (FSAR S 15A.1.4).

Other events are precluded by definition.

Thus, events such as con-l l

tinuous control rod withdrawal during power range operation l

(FSAR 5 15A.1.11) and operation of a fuel assembly in an im-proper location (FSAR $ 15A.1.16) cannot be postulated.

s (10)

In addition to the 18 events which simply cannot occur, there are 5 events for which the component operation evaluate'd in Chapter 15 could occur, but the phenomena of in-terest in Chapter 15 could not exist.

All recirculation pump events, such as recirculation pump trip (FSAR $ 15A.1.20) and abnormal startup of an idle recirculation pump (FSAR S 15A.1.25), would be of interest only if they could affect core physics or thermal-hydraulic conditions.

With no heat generation or boiling in the core, there are no pertinent phe-nomena (such as temperature differences or void collapses) to evaluate.

Another example, the core coolant temperature in-crease event (FSAR $ 15.A.1.26), postulates a loss of RHR cool-ing.

Even if the RHR system was operated in Phase I, there would be no temperature increase from decay heat to evaluate should the RHR system be lost.

(11)

The remaining 15 events addressed in Chapter 15 could possibly occur.

However, our review established that all are trivial events which have no potential to impact public health and safety.

Prior to initial criticality,.there are no fission products in the core and no decay heat exists.

It fol-lows that core cooling is not required.

In addition, with no fission product inventory, there are no fission product re-l leases possible.

Thus, for reactor events such as a control l

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rod removal error (FSAR 5 15A.1.13) and a control rod drop (FSAR S 15.1.33) and for non-reactor events such as a fuel han-dling abeident (FSAR 5 15.1.36) or a liquid radwaste tank rup-ture (FSAR~5 15.1.32), there could be no radiological conse-quences.

Therefore, there is no risk to public health and safety.

(12)

Even a loss of coolant accident (FSAR 5 15.1.34) could have no radiological consequences during Phase I.

No core cooling is required.

No fission product release is possi-ble.

The fuel simply could not be challenged by a complete draindown of the reactor vessel for an unlimited period of time.

(13)

In summary, the review of Chapter 15 events for fuel loading and precriticality testing indicates that many Chapter 15 events simply cannot occur, and for those that can, there can be no radiological consequences.

Therefore, there is no possible risk to the public health and safety.

This conclu-sion is not affected by any postulated diesel generator unavailability because it is in no way dependent on the avail-ability or unavailability of any AC power.

Phase II:

Cold Criticality Testing (14)

This phase of low power testing of the Shoreham

plant will include cold criticality testing of the plant at es-sentially ambient temperature and atmospheric pressure.

The power level during this phase of testing will be in the range of.0001% to.001% of rated power.

Details of the testing to be performed during this phase are described in the Notaro Af-fidavit.

(15)

The review of Chapter 15 revealed that of the 38 accident or transient events included there, 15 of the events could not occur because of the operating conditions of the plant during Phase I.

See Exhibit 2.

A number of these events are not possible because the reactor will be at essentially ambient temperature and pressure and no steam will be gener-ated.

For example, the generator load rejection event (FSAR S 15A.1.1) could not occur during this testing phase because steam is needed to drive the main turbine generator to permit ccnnecting it to the LILCO transmission system.

Another exam-ple, the loss of condenser vacuum event (FSAR 5 15A.1.21),

could not occur because it assumes that steam is available to draw a vacuum in the main condenser.

A third example, the in-advertent HPCI pump start event (FSAR S 15A.1.10), could not occur because there will be no steam available to power the HPCI pump, a steam driven ECCS pump.

Other Chapter 15 events could not occur because they are precluded by the configuration I

i i

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of the plant during this phase of low power testing.

An exam-ple of this type of event is the MSIV closure (FSAR $ 15A.1.4).

The MSIds will normally be closed throughout all of the op-erations conducted during this phase of low power testing.

In any event, there is no steam generated by the reactor to flow through the steam lines.

(16)

In addition to the 15 events that could not occur during Phase I, many of the 23 events remaining in the Chapter 15 analysis are far less likely to occur during low power testing than during normal operations.

For example, the recirculation pump trip (FSAR 9 15A.1.20), the recirculation pump seizure (FSAR S 15A.1.22), the recirculation flow control failures (FSAR $ 15A.1.23 and 24) and the abnormal startup of idle recirculation pump (FSAR $ 15A.1.25) events, although physically possible, are not as likely to occur because the recirculation pumps are used for only limited periods of time during this phase of the testing program.

Similarly, the loss of feedwater event (FSAR $ 15A.1.18) is very unlikely because little, if any, make-up water will have to be supplied to the reactor.

Moreover, make-up water would not normally be supplied by the feedwater system under these conditions.

Other-very unlikely events include miscellaneous small releases out-side primary containment (FSAR 5 15.1.29), off design

t i '

operational transient as a consequence of instrument line fail-ure (FSAR $ 15.1.30), and feedwater system piping break (FSAR

$ 15.1.37).

Thus, many of the Chapter 15 events that are phys-ically possible during Phase II remain very unlikely in light 4

of the plant conditions that will then exist.

I (17)

Nonetheless, all 23 possible events contained in Chapter 15 were reviewed to reaffirm that the consequences of these events, should one occur during Phase I of low power testing, would be bounded by the consequences analyzed for the event considered in the FSAR.

A discussion of some of the 23 l

l possible events contained in Chapter 15 illustrates the basis for this conclusion.

The continuous control rod withdrawal during startup event (FSAR $ 15A.1.12) is applicable to op-eration in the power, source and/or intermediate range of op-eration.

During cold functional criticality testing, the reac-tor will operate in the source and intermediate ranges and therefore the conclusions contained in Chapter 15 are applica-j ble to this event should it occur during this phase of low power testing.

As the-FSAR indicates, this event would not re-sult in any release of radioactive material from the fuel at any power level.

-Another example is the fuel handling accident j

(FSAR $ 15.1.36).

As stated in the FSAR, the_most severe fuel handling accident from a radiological viewpoint is a dropping b

.,-ee g

y

~-

of the fuel assembly onto the top of the core.

The ESAR analy-sis assumes that the fuel contains a fission product inventory equivaleht to operation of 1000 days at full rated power.

This assumption results in an equilibrium fission product concentra-tion at the time the reactor is shut down.

But as already noted, the fission product inventories in the core will be sig-nificantly less during Phase II low power testing than the in-ventories analyzed in the FSAR because of the extremely low power levels (.0001% to.001% of rated power) achieved during this testing.

Thus, even if a handling accident took place and fuel damage did occur, there would be significantly less fis-sion products to be released from the core than those that have already been analyzed and found accept'able in the FSAR.

A third example is the liquid radwaste tank rupture event (FSAR $

15.1.32).

This event assumes the rupture of a radwaste tank that contains a substantial amount of contaminated liquids gen-erated during the operation of the reactor.

But again, since Phase II low power testing results in insignificant power lev-els in the reactor, there will be little, if any, radioactive liquids in the radwaste tank should such a rupture occur.

Thus, even the minimal consequences already described in the FSAR for the design basis event would be further reduced under these low power testing conditions.

For each of these events, the review concluded that the consequences are significantly

less severe for any event occurring during the cold functional criticality testing than for the event analyzed in Chapter 15.

To summatize, because of the extremely low power levels reached during this' testing phase, fission product inventory in the core will be only a small fraction of that assumed for the Chapter 15 analyses.

As indicated above, the FSAR assumes op-eration at 100% power for 1000 days in calculating fission product inventory; the inventory during Phase II low power testing will be less than one one-hundred-thousandth (.00001) of the fission product inventory assumed in the FSAR.

Conse-quently, none of the events analyzed in Chapter 15 could result in a release of radioactivity during cold criticality testing that would harm the public health and safety.

(18)

The review of Chapter 15 events for Phase II testing and the conclusions reached are unaffected by any unavailability of the TDI diesels.

Of the 23 possible Chapter 15 events reviewed, 20 of the events in the FSAR do not require the assumption of loss or unavailability of offsite AC power.

See Exhibit 2.

Thus, our conclusions for these 20 of the 23 possible events are independent of the status of the diesels.

I I

(19)

The three events'that do assume loss or unavailability of offsite AC power are (1) pipe breaks inside the primary containment (LOCA) (FSAR $ 15.1.34), (2) feedwater l

l l

system piping break (FSAR $ 15.1.37), and (3) the loss of AC power event (FSAR S 15A.1.19).

With respect to these events, the LOCI'would be the most limiting event.

The review has shown that'if a LOCA did occur during the cold criticality testing phase, however remote that possibility, there would be time on the order of months available to restore make-up water for core cooling.

At the power levels achieved during Phase II, fission product inventory is very low.

At most, decay heat will, on the average, be a fraction of a watt per rod, with no single rod exceeding approximately 2 watts.

This is less, roughly, than the heat output of a Christmas tree bulb.

It follows that the fuel cladding temperature would not exceed the limits of 10 CFR $ 50.46 and Appendix K even after months with-out cooling and without any source of AC power.

(20)

The loss of AC power event (FSAR S 15A.1.19) and the feedwater system piping break (FSAR S 15.1.37) under cold criticality testing conditions do not rely on the diesel gener-ators for mitigation of the event.

For these events, since no loss of coolant occurs and the decay heat is minimal, core cooling is achieved, without AC power, using the existing core water inventory and heat losses to ambient, for essentially un-limited periods of time.

In'any event, as demonstrated in the l

Schiffmacher Affidavit, AC power sources can and will be i

I

readily supplied to the Shoreham plant even if one assumes the simultaneous loss of all three emergency diesel generators.

(21)

In addition to our conclusions that the limiting LOCA event could not approach the limits of 10 CFR 6 50.46 and Appendix K during Phase II low power testing, there are other reasons why our findings with respect to the three events that assume loss of AC power are independent of the availability of the TDI diesels.

The LOCA (pipe break inside containment) and the feedwater system piping break postulate the double ended rupture of a piping system.

Because the reactor will be at es-sentially ambient temperature and atmospheric pressure during Phase II, it is extremely unlikely that such a pipe break would ever occur.

In fact, the NRC Staff does not require double ended ruptures to be postulated for low temperature and low pressure systems in safety analyses.

Thus, these events are much less likely during cold criticality testing than during normal operation.

i (22)

The review of Chapter 15 events for cold criti-cality testing indicates that performance of these activities at Shoreham involves essentially no risk to the public health and safety.

This conclusion is not affected by any postulated r

diesel unavailability.

In fact, even if AC power were not available for extended periods of time, fuel design limits and I

~

l 17-design conditions of the reactor coolant pressure boundary would not be approached or exceeded as a result of anticipated operatihnaloccurrences, and the core would be adequately cooled in the unlikely event of a postulated accident.

Phases III and IV:

Low Power Testing Up To 5% of Rated Power (23)

These aspects of low power testing will include operation of the plant at power levels up to 5% of rated power.

Details of the testing to be performed during this phase of op-eration are described as Phases III and IV in the Notaro Affi-davit.

(24)

The review of the 38 Chapter 15 events for these phases of low power testing operations revealed that two of the events in Chapter 15, generator load rejection-(FSAR 6 15A.1.1) and turbine trip with generator breaker failure (FSAR 6 15.1.2) cannot occur because the generator will not be connected to the-grid during these phases of testing.

A third event,.the cask drop, is precluded by desiga as stated in FSAR 5 15.1.28.

See Exhibit 3.

(25) of the remaining 35 events that can occur during this phase of operation, 31 of the events do not assume loss or 4

unavailability of AC power.

For each of these 31 events, 4

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operation of the plant up to 5% of rated power will be bounded by the Chapter 15 analysis.

Since the Chapter 15 analysis con-siders all possible phases of plant operation, it follows that operation at 5% can result in consequences less severe than those analyzed in Chapter 15.

For example, the turbine trip event (FSAR $ 15A.1.2) assumes that the limiting event occurs with the reactor operating at 105% of rated steam flow coupled with failure of the turbine bypass valves to open.

Even this limiting event does not result in any fuel failures.

FSAR S 15A.1.2 specifically notes that turbine trips at power levels less than 30% of rated power are bounded by the limiting analy-sis.

Another example is the loss of feedwater heating event (FSAR S 15A.1.8).

This event assumes continuous operation of the feedwater system and the most severe possible loss of feedwater heating, resulting in the injection of colder feedwater.

For operation at power levels less than 5%, the im-pact of lost feedwater heating is minimal because of the low feedwater flow.

Since these analyses are not required to as-sume the absence of AC power, potential unavailability of the TDI diesels has no effect on the assessment of these events.

(26)

Not only are the results of these 31 events bounded by the Chapter 15 analysis, the consequences of these events are also less than the consequences stated in the FSAR.

l

First, the power limitations during low power testing up to 5%

power, the fission product inventory in the core will not ex-b ceed 5% of the values assumed in the FSAR.

In fact, because of the intermittant type of operations conducted during low power testing, equilibrium fission product inventory for even 5%

power is unlikely to be achieved.

This low fission product in-ventory reduces risk in two ways:

(a) the amount of decay heat present in the core following shutdown is substantially re-duced, and (b) the amount of radioactivity that could be re-leased upon fuel failure is substantially reduced.

(27)

The second factor contributing to the signifi-cantly lower risk during low power operation is the increased time available for preventive or mitigating action should such action be deemed desirable by the operator.

Longer time is available because the limited power levels mean that it takes longer for the plant to reach setpoints and limits.

For exam-ple, on loss of feedwater (FSAR S 15A.1.18), the water level in the reactor will decrease at a slower rate than if the event occured at 100% power.

This gives the operator more time to act manually to restore feedwater before an automatic action takes place.

Similarly, in the loss of condenser vacuum event (FSAR 5 15.A.1.21), the operator will have more time to identi-fy the decreasing vacuum and to take steps to remedy the

situation before automatic actions such as turbine trip, feedpump trip or main steam isolation occur.

Another example is the mhin steam isolation valve closure event (FSAR S 15A.l.4).'

At five percent power, the amount of heat produced upon isolation of the reactor vessel (which is followed by a reactor scram) results in a much slower pressure and tempera-ture increase than would be experienced at 100% power.

This gives the operator more time to manually initiate reactor cool-ing rather than relying on automatic action.

In effect, the operator may end the transient before there is any substantial impact on the plant.

(28)

The third factor contributing to the signifi-cantly lower risk during low power testing is the reduction in the required capacity for mitigating systems.

Because of the lower levels of decay heat present following operation at 5%

power, the demand for core cooling and auxiliary systems is substantially reduced, permitting the operation of fewer sys-tems and components to mitigate any event.

It follows that the AC power requirements for event mitigation are substantially reduced for 5% power operation as compared to 100% power op-eration.

(29)

As already noted, only four of the events ana-lyzed in Chapter 15 require the assumption of the 1

t

unavailability of offsite AC power for operation during Phases III and IV.

Of these four events, the loss of coolant accident is the ost limiting event.

The Chapter 15 LOCA analysis as-sumes the unavailability of offsite AC power.

This is a con-servative licensing assumption.

In fact, as described in de-tail in the Schiffmacher Affidavit, there are multiple sources of AC power available to the Shoreham site (e.g., emergency diesel generators, two normal sources of offsite power, blackstart gas turbines at Holtsville, Southhold, and East Hampton, a blackstart gas turbine on the Shoreham site, and mo-bile diesel generators).

Thus, AC power will be available at Shoreham to mitigate a loss of coolant accident during low power operations up to 5% rated power.

In the unlikely event offsite AC power is lost, it can be restored within sufficient time to prevent exceeding the limits of 10 CFR $ 50.46 and Ap-pendix K.

GE has determined that for 5% power so long as reflooding of the core has occurred within approximately one hour, S 50.46 criteria will be met.2/ As the Schiffmacher Af-fidavit demonstrates, power can be restored to Shoreham within minutes.

An evaluation has been performed to assure the ade-quacy of containment isolation in the event AC power sources 2/

As shown in the Exhibit 4 below, lower power levels will result in more time to restore power and core cooling for a postulated LOCA.

Thus, for 1% power approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> are available.

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l

cannot provide immediate isolation in a LOCA.

Based upon the results of this evaluation, we have concluded that through the j

use of $~ppropriate manual action, containment isolation can be accomplishbd in a timely manner.

1 (30)

For the other three events, (1) loss of AC power (FSAR $ 15A.1.19), (2) pipe break outside containment (PBOC)

(steam line break accident) (FSAR $ 15.1.35) and (3) feedwater system piping break (FSAR $ 15.1.37), the reactor would auto-matically isolate.

This isolation is not dependent upon the availability of AC power.

For all three events, both HPCI and RCIC would be available to provide reactor coolant makeup.

Given the heat capacity of passive heat sinks such as structur-al steel, suppression pool cooling would not be required for about 30 days.

Therefore, there is ample time for AC power to be restored.

Furthermore, assuming loss of offsite power in the context of pipe breaks outside containment (main steam line break accident and feedwater system break accident) is a con-servatism which stems from the PBOC analysis methodology.

That methodology requires the assumption of a loss of offsite power for pipe breaks which result directly in a plant trip of the turbine generator system or reactor protection system.

Not-withstanding grid stability analyses, it is assumed that plant trips could cause perturbations of the grid, resulting in the

)

loss of offsite power.

For operation at 5% power or less, how-ever, the turbine generator is not connected to the grid, and thereforb any assumption of induced perturbation to the offsite grid is not valid.

(31)

Based on our review of Chapter 15, operation of the plant during low power testing up to levels of 5% of rated power poses no undue risk to the public health and safety.

In fact, any risk is substantially less than that already found to be acceptable by the NRC Staff in its review of Chapter 15.

Even if the Shoreham TDI diesels are assumed to be unavailable, there is ample assurance that fuel design limits and design conditions of the reactor coolant pressure boundary will not be exceeded as a result of anticipated operational occurrences, and that the core will be cooled and containment integrity and other vital functions will be maintained in the event of any postulated accident.

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Eugey& C.

Eckert

STATE OF Maryland _)

)

To-wit:

COUNTY OF Montgomery)

Subscribed to before me this 29tilay of March, 1984.

t as to Glenn G. Sherwood and Eugene C. Tckert ns) OD.

Dr

/ ren M. ThompsonNotary Publicf Ka My commission expires:

July 1, 1986 1

i i

l i

Exhibit 1 FUEL LOAD AND PRECRITICALITY TESTING Chapter 15 Event Event Category 1.

Generator Load Rejection 2.

Turbine Trip 3.

Turbine Trip with Failure of Generator Breakers to Open 4.

MSIV Closure 5.

Pressure Regulator Failure - Open 6.

Pressure Regulator Failure - Closed 7.

Feedwater Controller Failure -

Maximum Demand 8.

Loss of Feedwater Heating 9.

Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature 10.

Inadvertent HPCI Pump Start 11.

Continuous Control Rod Withdrawal During Power Range Operation Event not possible.

Component operation possible but Chapter 15 phenomena cannot occur.

Event possible but no consequences.

l

12.

Continuous Rod Withdrawal During Reactor Startup 13.

Control Rod Removal Error During Refueling 14.

Fuel' Assembly Insertion Error During Refueling 15.

Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location 16.

Inadvertent Loading and Operation of a Fuel Assembly in Improper Location 17.

Inadvertent Opening of a Safety / Relief Valve 18.

Loss of Feedwater Flow 19.

Loss of AC Power 20.

Recirculatien Pump Trip 21.

Loss of Condenser Vacuum 22.

Recirculation Pump Seizure 23.

Recirculation Flow Control Failure -

Decreasing Flow 24.

Recirculation Flow Control Failure With Increasing Flow 25.

Abnormal Startup of Idle Recirculation Pump l

26.

Core Coolant Temperature Increcse 27.

Anticipated Transients Without SCRAM (ATWS) 28.

Cask Drop Accident l

29.

Miscellaneous Small Releases Outside Primary Containment t

30.

Off Design Operational Transient as a Consequence of Instrument Line Failure 31.

Maln Condenser Gas Treatment System Failure 32.

Liquid Radwaste Tank Rupture 33.

Control Rod Drop Accident 34.

Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident) 35.

Pipe Breaks Outside Primary Containment (Steam Line Break Accident) 36.

Fuel Handling Accident 37.

Feedwater System Piping Break 38.

Failure of Air Ejector Lines i

l l

Exhibit 2 COLD CRITICALITY TESTING Assumes Un-Event availability Chapter 15 Event Category of Offsite AC l

1.

Generator Load Rejection N/A 2.

Turbine Trip N/A 3.

Turbine Trip with Failure of Generator Breakers to Open N/A 4.

MSIV Closures N/A 5.

Pressure Regulator Failure - Open N/A 6.

Pressure Regulator Failure - Closed N/A 7.

Feedwater Controller Failure -

Maximum Demand No 8.

Loss of Feedwater Heating N/A 9.

Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature No 10.

Inadvertent HPCI Pump Start N/A 11.

Continuous Control Rod Withdrawal During Power Range Operation N/A Event not possible.

Event possible but essentially no consequences.

i 4

12.

Continuous Rod Withdrawal During Reactor Startup No 13.

Control Rod Removal Error During Re' fueling No 14.

Fuel' Assembly Insertion Error During Refueling No 15.

Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly into an Improper Location No 16.

Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.

Inadvertent Opening of a Safety / Relief Valve N/A 18.

Loss of Feedwater Flow No 19.

Loss of AC Power Yes 20.

Recirculation Pump Trip No 21.

Loss of Condenser Vacuum N/A 22.

Recirculation Pump Seizure No 23.

Recirculation Flow Control Failure -

Decreasing Flow No 24.

Recirculation Flow Control Failure With Increasing Flow No 25.

Abnormal Startup of Idle Recirculation Pump No 26.

Core Coolant Temperature Increase No 27.

Anticipated Transients Without SCRAM (ATWS)

No 28.

Cask Drop Accident N/A 29.

Miscellaneous Small Releases Outside Primary Containment No l

l

30.

Off Design Operational Transient as a Consequence of Instrument Line Failure No 31.

Mk'inCondenserGasTreatment System Failure N/A 32.

Liquid Radwate Tank Rupture No 33.

Control Rod Drop Accident No 34.

Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)

Yes 35.

Pipe Breaks Outside Primary Containment (steam line break accident)

  • N/A 36.

Fuel Handling Accident No 37.

Feedwater System Piping Break Yes 38.

Failure of Air Ejector Lines N/A

Exhibit 3 5% POWER Assumes Un-Event availability Chapter 15 Event Category of Offsite AC 1

1.

Generator Load Rejection N/A 2.

Turbine Trip No 3.

Turbine Trip with Failure of Generator Breakers to Open N/A 4.

MSIV Closures No 5.

Pressure Regulator Failure - Open No 6.

Pressure Regulator Failure - Closed No 7.

Feedwater Controller Failure -

Maximum Demand No 8.

Loss of Feedwater Heating No 9.

Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature No 10.

Inadvertent HPCI Pump Start No 11.

Continuous Control Rod Withdrawal During Power Range Operation No Event cannot occur.

Bounded by same event at higher power level per FSAR Chapter 15.

i.

I 12.

Continuous Rod Withdrawal During Reactor Startup No 13.

Control Rod Removal Error During Refueling No 14.

Fuel' Assembly Insertion Error During Refueling No 15.

Off-Design Operational Transients Due to Inadvertent Loading of a Fuel Assembly Into an Improper Location No 16.

Inadvertent Loading and Operation of a Fuel Assembly in Improper Location No 17.

Inadvertent Opening of a Safety / Relief Valve No 18.

Loss of Feedwater Flow No 19.

Loss of AC Power Yes 20.

Recirculation Pump Trip No 21.

Loss of Condenser Vacuum No 22.

Recirculation Pump Seizure No 23.

Recirculation Flow Control Failure -

Decreasing Flow No 24.

Recirculation Flow Control Failure -

l With Increasing Flow No 25.

Abnormal Startup of Idle Recirculation Pump No 26.

Core Coolant Temperature Increase No 27.

Anticipated Transients Without SCRAM (ATWS)

No 28.

Cask Drop Accident N/A t

29.

Miscellaneous Small Releases Outside Primary Containment No t

l l

l

30.

Off Design Operational Transient as a Consequence of Instrument Line Failure No 31.

Md'in Condenser Gas Treatment System Failure No 32.

Liquid Radwaste Tank Rupture No 33.

Control Rod Drop Accident No 34.

Pipe Breaks Inside the Primary Containment (Loss of Coolant Accident)

Yes 35.

Pipe Breaks Outside Primary Containment (Steam Line Break Accident) **

Yes 36.

Fuel Handling Accident No 37.

Feedwater System Piping Break Yes 38.

Failure of Air Ejector Lines No

o ECCS LOCA EVALUATIONS 10 CFR E 50.46 Limits Co re Peak Rod Time to 10 PCT Local Core Wide Avg. Powe r MAPLHCR CFR i 50.46 (F')

Oxidation Oxidation (1 or rated)

(kW/Ft1 Limits (mini flimit 2200')

(Limit 171)

(Limit 111 5.0 1.34 55 2200 6.5 less than 0.9 2.5 0.67 124 2200 8.4 less than 1.0 1.25 0.34 285 2100 9.0 1.0

.5 0.13 700 2000 9.0 1.0 ASSUMPTIONS: 10 CFR 50 Appendix K (Standard FSAR Basis)

Initial Conditions Based on Equivalent Core at Designated Core Average Power a

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