ML20206T769

Forwards Rev 17 to Radiation Emergency Plan.Encl Addresses Three Inconsistent Items Identified in NRC Re Classification of Emergencies

ML20206T769
Person / Time
Site:	Summer
Issue date:	06/27/1986
From:	Nauman D SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML20206T773	List: ML20206T769
References
NUDOCS 8607080223
Download: ML20206T769 (7)

Text

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So th ro na Electric & Gts Company n .Nw n Columbia SC 29218 Nuclear Operatons (B03) MS-3513 SCE&G M. 8. _

t June 27,1986 Dr. J. Nelson Grace Regional Administrator U.S. Nuclear Regulatory Commission Region ll, Suite 2900 101 Marietta Street, N.W.

Atlanta, Georgia 30323

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Emergency Plan Review

Dear Dr. Grace:

This letter is provided in response to NRC letter dated May 16,1986, on the Virgil C.

Summer Nuclear Station Radiological Emergency Plan Review. Enclosed are two (2) copies i of Revision 17 to the Radiological Emergen,cy Plan (REP) which includes changes in emergency action levels, emergency organization, emergency personnel exposure and protection action recommendations during a General Emergency.

The enclosure to the May 16 letter specifically identified three items of " inconsistencies in the Plan" involving the classification of emergencies. Atthe presenttime,the inconsistencies have been addressed in Revision 17. South Carolina Electric & Gas has a number of questions and concerns on the bases for which the NRC determined the inconsistencies in the plan. The attached Enclosure addresses each specific item with additional information and bases for the emergency classifications as proposed and is provided to assist in the final review of Revision 15 to the REP.

This revision, which has been reviewed by the Plant Safety Review Committee, does not decrease the effectiveness of the Radiological Emergency Plan to meet the standards of 10CFR50.47(b) and Appendix E.

If you have any questions, please notify this office.

Vqry truly yours, Q-

. s

- .s s s x . + . .

D. A. Naumah ,

A 8607080223 860627 PDR ADOCK 05000395

{( M\ '

F PDR

. , l Dr. J. Nelson Grace Page Two June 27,1986 c: O. W. Dixon, Jr./T. C. Nichols, Jr.

E. H. Crews, Jr.

E. C. Roberts W. A. Williams, Jr.

Group Managers O. 5. Bradham D. R. Moore C. A. Price S.R. Hunt C. L. Ligon (NSRC) -

K. E. Nodland R. A.Stough 4

G. O. Percival R. L. Prevatte J. B. Knotts, Jr.

NPCF File 4

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Page 1 ENCLOSURE

1. Section 4.0, Table 4-1, Pac'e 29: As an initiating condition, the loss i of all annunciator alarms eading to the Notification of Unusual Event (NOUE)is inconsistent with NUREG-0654, Appendix 1, Pag"e 1-9, item 14, which states "most or all alarms (annunciators) lost -

as an initiating condition for the Alert classification. J RESPONSE: The class description of an Alert as stated in NUREG-0654, Revision i 1, asserts that an Alert is a condition involving an actual or potential -

substantial degradation of the level of safety of the plant. The  : -

annunciator system is classified as a non-nuclear safety system as evidenced by its absence from Table 7.1.2 of the FSAR. The failure '

of non-nuclear safety class components must be incorporated into -=

the design considerations for the plant so as to preclude a '

compromise of plant safety. By design, non-nuclear safety

~

_ ' components cannot compromise safety. If a system cannot affect Z safety,it cannot result in a actual or potential substantial -

degradation of level of safety and its failure or loss would not 3 correspond to the class description for an Alert in NUREG-0654,

• Revision 1.

2. Section 4.0, Table 4-1, Page 29: The NOUE classification based on the loss of all onsite D-C power for a period greater than 5 minutes, ]

is inconsistent with NUREG-0654, Appendix 1, Page 1-9, item 8, =

which states " loss of all onsite DC power" as an initiating condition for the Alert classification. _

RESPONSE: The class description of an Alert as stated in NUREG-0654, Revision 3 1, asserts that an Alert is a condition involving an actual or potential substantial degradation oflevel of safety of t1e plant. While the a total loss of all DC power in certain nuclear power plants may result -

in a substantial degradation of safety,in the case of the V. C. -

Summer Nuclear Station safety would not be substantially -

degraded upon the loss of all DC power. The loss of all DC 3ower -

was ofinterest to the ACRS during its evaluation associatec with 2 the issuance of an Operating License for the V. C. Summer Nuclear Station. The loss of DC power results in the transfer of

• instrumentation power sources to vital AC sources. The ability to  :

remotely control intermediate voltage (7.2 kv) switchgear J supplying internal Station buses would be lost; however, the -

intermediate voltage switchgear would remain as is upon the loss J of DC power. The short term loss of DC power does not cause i conditions which result in the release of radioactive material to the ..

environment which could represent a hazard to the public health i and safety.

]4 The loss of all onsite DC power will not result in the loss of all AC 9 power. DC pow ~er is necessary to effect remote control over large switchgear. The breakers associated with the 7.2 kv and 480 V buses within the Station are DC controlled. The remote and -

automatic operation of the 7.2 kv and 480 V breakers is dependent "

upon availability of DC control power. Manual operation of the 7.2 '-

kv and 480 V breakers is independent of the status of DC control  ;

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Page 2 i.

power. The low voltage motor control centers are AC controlled.

The source of AC control power for low voltage motor control center is the AC supply bus for the motor control center. With AC power available, large pumps can be controlled manually at the switchgear. Small pumps and AC powered remotely controlled valves will still be controlled by AC power. Certain control valves, such as the turbine driven auxiliary feedwater pump supply and

, discharge valves, are DC controlled with local manual operation 4

capability. If operation of those valves is necessary, local manual

, operation will be implemented. Necessary instrumentation is powered by either vital AC,'which will remain energized, or

' inverted DC which will be unavailable. The loss of a 48 VDC output from the Solid State Protection System will result in a reactor trip due to the use of a DC undervoltage trip mechanism in the breaker supplying rod drive control power. However,the Solid State protection system is powered from inverters '

and would not lose power from loss of the station batteries.

The loss of all onsite DC power may complicate the recovery from an automatic reactor trip. Nevertheless, vital services could continue i

or be reestablished by means of operator action since offsite AC -

power will remain available. The means to provide auxiliary feedwater, charging, and necessary cooling to auxiliaries will exist.

The loss of DC power does not directly result in the loss of any fission product barrier. The initial classification of the loss of all onsite DC power would be Notification of Unusual Event.

Subsequent degradation of any fission product barriers would be

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observable by the operators in sufficient time to implement an appropriate emergency response. In the worst case, the total loss of i DC power corresponds to the total loss of all AC power in terms of time to breach of fission product barriers and development of conditions which present a hazard to the public. The premature declaration of an Alert, Site Area, or General Emergency condition i is not warranted due to the substantial time between loss of DC

power, and development of conditions which could result in significant release of radioactive material to the environment so as to present a hazard to the health and safety of the public. The classification of this event employing the Fission Product Barrier Approach corres 3onds to the level of actual hazard which should be maintained at a ow level by the effective mitigation activities of the operators.

3. Section 4.0, Table 4-3, Page 32F:'The initiating condition for the .

l. alert classification involvmg the loss of offsite power and loss of all pnsite AC power was not included in this table, This change is inconsistent with NUREG 0654, Appendix 1, Page 1-9, item 7.

RESPONSE: The design of the V. C. Summer Nuclear Station includes a turbine-

, driven auxiliary feedwater pump which is sized to provide a

sufficient water supply to the steam generators should all AC power be lost. The loss of all AC power will not immediately result in the degradt. tion of any fission product barrier. The loss of all AC power is considered a Notification of Unusual Event.

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Page 3 The loss of both onsite and offsite AC power does not directly result in the release of radioactive material, nor the breach of any fission product barriers. As stated above, a turbine driven auxiliary feedwater pump is available to su oply auxiliary feedwater to the steam generators, independent of the availability of AC power. The i

' auxiliary feedwater pump assures that an adequate means of decay heat removal exists by steam relief from the steam generators to atmosphere. The turbine driven auxiliary feedwater pump, steam supply valves, and condensate supply valves are DC powered and  ;

remain functional u,3on the loss of all AC. Additionally, the turbine driven auxiliary feec! water pump and its associated supply and discharge valves may be manually controlled by operators to assure

, adec uate heat removal. The Westinghouse designed NSSS emp oyed at the V. C. Summer Nuclear Station includes a sufficient l water mventory in the steam generators to provide decay heat I 4

removal for approximately 25 minutes following the loss of all AC power. Sufficient time exists for the operators to establish auxiliary ,

feedwater by means of the turbine driven pump prior to the loss of- l

steam pressure due to steam generatordryout. A sufficient supply '

of water to provide adequate decay heat removal is assured due to Technical Specification requirements for condensate storage tank inventory.

During the initial stages of a total loss of AC power no fission product barrierwould be breached. Any degradation of any fission product barrier will be indicated on plant instruments since

)

instrumentation to accomplish fission product barrier status assessment is DC powered or supplied by vitalinstrumentation bus 4

l AC which is connected to the DC supply by means of inverters.

i l The first fission product barrier subjected to degradation due to the l

5 loss of AC power would be the reactor coolant system pressure boundary. The likely means to result in breach of the reactor coolant system pressure boundary fission product barrier is the 1

degradation of a reactor coolant pump seal or steam degradation

of a reactor coolant pump seal or steam relief fr'om the reactor i

coolant system due to loss of decay heat removal capability by the

, steam generators, resulting from the loss of water inventory in the generators.

In the event that the reactor coolant system begins to lose mass due either to a pump seal failure or steam relief to tae Reactor Building through the pressurizer safety valves, the inventory of the reactor coolant system is adequate to provide core cooling for a significant period of time. Once continuous mass loss from the reactor coolant i ,

system has begun, unless means are provided to replace the reactor i

coolant at a rate equal to or greater than the boil-off rate of the j

core in a pool boiling mode of heat transfer, core damage will result when insufficient inventory of coolant exists within the core to accomplish adequate cooling. Upon the loss of all AC power -

r combined with the unavailability of any means to add water to the steam generators or the reactor coolant system, a number of hours of heat removal capability exists due to the water inventory within

_ _ ._m____ _ , _ _ _ . _ _ __ _ _ _ . _ _. _. . . _ . _ , . __

Page 4 the steam generators and the reactor coolant system. If the loss of mass from the reactor coolant system is by means of the aump seals

- with auxiliary feedwater available, the time to onset of fuel damage may be days.

' i

)- Immediately following the loss of AC power, no fission product barriers would be breached. Subsequent degradation is possible, warranting continuous surveillance for indications of the onset of 1 fission product barrier breach. The loss of AC power corresponds to a Notification of Unusual Event classification as long as all fission product barriers remain functional. Reclassification to a higher level of emergency would occur with the loss of function of any fission product barrier. The minimum time for the development of such symptoms would be approximately 30 minutes, assuming an absence of effective mitigation by operators. Subsequent fission -

product barrier failure will be dependent upon the effectiveness of

, mitigation activities and the duration of de AC power loss. Since

< hours of core cooling capacity are available, there is no need to prematurely upgrade the event classification to assure public safety. Throughout the event the indictions of fission product barrier status would enable the effective contemporaneous 4

assessment of the level of hazard such that the event classification would be consistent with the actual level of hazard to the public.

The loss of all AC power was also of interestto the ACRS during its Operating License review. The presentations made by SCE&G to the ACRS identified specific design features of the V. C. Summer Nuclear Station which mitigate the loss of all AC power for a

number of hours during which time corrective measures to recover j AC power from multiple potential sources will be implemented.

j The NRC staff acknowledged the existence of these design features i and agreed with SCE&G's conclusions during those ACRS review

meetings. The conclusion asserted to the ACRS was that the V. C.

Summer Nuclear Station includes specific design features which mitigate the totalloss of all AC power for a substantial period of i time such that a significant degradation of the level of safety would not exist. Considering the above, the loss of all AC power within

! the Summer Station is not a condition which corresponds to the i class description for an Alert. Subsequent events or degradation of l

' components could result in conditions which would correspond to l the criteria for an Alert as specified in NUREG-0654, Revision 1. The 1 l Fission Product Barrier Approach to Emergency Event Classification 3

includes provisions for properly assessing any subsequent events or

degradation so as to properly classik the actual and potential (unavoidable) degradation in level of safety in conformance to the i

guidance of NUREG-0654, Revision 1, as such conditions develop.'

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Je VIRGIL C. SUMMER NUCLEAR STATION RADIATION EMERGENCY PLAN

• REVISION 17 INSTRUCTION SHEET The following instructional information and checklist is being furnished to help insert Revision 17 into the Virgil C. Summer Nuclear Station Radiation Emergency Plan.

l Discard the old sheets and insert the new sheets as listed below. Keep these instructions in the front of the Radiation Emergency Plan to serve as a record of changes.

Remove Insert Title Page Title Page List of Tables List of Tables Page 11 Page 11 l

Page 13 Page 13 Page 29 Page 29 Page 32H Page 32H Table 4-4, Page 321 Page 34 Page 34 Page 43 Page 43 Page 47 Page 47 Page 49 Page 49 Figure 5-2 Figure 5-2 Figure 5-3 Figure 5-3 i

Page 58 Page 58 Table 6-3, Page 61 K Table 6-3, Page 61K Page 65 Page 65 Page 66 Page 66

, Page 78 Page 78  ;

Figure 7-2 Figure 7-2 l

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