ML19031A947
| ML19031A947 | |
| Person / Time | |
|---|---|
| Site: | Salem  (DPR-070, DPR-075) |
| Issue date: | 07/30/1976 |
| From: | Mittl R Public Service Electric & Gas Co |
| To: | Kniel K Office of Nuclear Reactor Regulation |
| References | |
| Download: ML19031A947 (5) | |
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PS~G e Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /622-7000
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July 30, 1976 Director of Nuclear Reactor Regulation u.s.. Nuclear Regulatory Commission Washington, D. C.
20555 Attention*:
Mr. Karl Kniel, Light Water Reactors Gentlemen:
REQUEST FOR DEFERMENT INCOMPLETE ITEMS NO., 1 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-272 Supplementing our *letters of July 14, and July 20, 1976, PSE&G hereby transmits additional items for which deferment until after core load is requested,;
The attachment to this letter identifies additional items which will be completed prior to initial criticality along with our safety evaluation.
We have concluded that the items identified in the attachment do not involve an unreviewed safety question and will have no effect on the safe operation and reliability of the plant.
The Energy People Very truly yours, R. L. Mittl General Manager - Projects Engineering and Construction Department 95-2001 (400M) 5-73
ITEM Axial Flux Difference Monitor (AFDM)
Radiochemistry Procedures ITEMS TO BE COMPLETED PRIOR TO INITIAL CRITICALITY DESCRIPTION The AFDM detects the difference in neutron flux between the top and bottom halves of the core.
.1 Procedure #PD 3.3.010
.2 Procedure #PD 3.3.011
- 3 Procedure #PD 3.3.003 SAFETY EVALUATION The Technical Specifications require control of axial flux difference during power. operations.
~'7he monitor will be installed ~n the control room at the time of core load and will be calibrated and tested prior to initial criticalit&
The device performs no function..
until power operation.. Neutron* flux is measured by other devices during initial core load.
Not having this device calibrated at the time of core load will not affect the control or evaluation of reactor conditions and therefore will have no adverse effect on the health and safety of the public.
.1 Procedure to determine the average energy (E.) of gamma emitting isotopes
- This procedure is required fo::9' normal operations (Mode 1) and relates to the builaup of fission products as the result of operation.
The Tec1mical Specifications (Table 4.4-4) require that the sample be taken after a minimum of 2 El'PD and 20 days of power Operation.
Radiochemistry procedures for control of the Reactor Coolant
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ITEM ITEMS TO BE COMPLETED PRIOR TO INITIAL CRITICALITY DESCRIPTION SAFETY EVALUATION
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System have been approved and are available for core load.
The lack of this procedur,~ at core load will not affect the capability to evaluate Reactor Coolant conditions. and therefore, will have no adverse A.
effect on the health and
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safety of the public.
This procedure will therefore :..)e completed prior to initia.L criticality *
- 2 Procedure for detecting fission gases by gamma spectrosco-:?Y in the presence of other g*ases.
This procedure is used follow-ing power operations to identify fuel element leakage.
Otner radiochemistry procedures are.
approved and availablE~ fo:c core load to identify isotopes and elements in the reactor coolant.
The lack of this procedure at core load will not affect the capability to evaluate reactor coolant conditions, since fission gases are not produced until critical-ity is achieved and there.fore, will have no adverse affect on the health and safety of ~he public.
This procedure will be completed prio'.!'." to initial criticality.
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ITEM Auxiliary Building Ventilation System Standby Charcoal Filter ITEMS TO BE COMPLETED PRIOR TO INITIAL CRITICALITY DESCRIPTION This charcoal filter removes 90"/o of the elemental and methyl (organic) iodines contacted at the rated flow of 21,400 cfm.
These iodines are those that become airborne in the auxiliary building due to leakage from the ECCS during recirculation after a design basis loss of coolant accident.
The filters are presently installed and have the capability to remove 90"/o.of the elemental iodine.
These filters dO not, however SAFETY EVALUATION
.3. Procedure to determine the Dose
- Equivalent Iodine 131 in the primary coolant.
This pro.cedure is used to equate the variety of iodine isotopes into a Dose Equivalent Iodine 131 value.
It relates to the A
"-buildup of fission gases as t119' result of power operation.
Radiochemistry Procedu::es for identifying all isotopes of Iodine are approved and avail-able for core load.
The lack of this procedure at core load will not affect the capabil.Lty to evaluate Reactor Coolant conditions since there is no buildup of I-131 until after initial*criticality is achieved and therefore, will have no affect on the health and safety of the public.
This p:cocedure
- will be ~ompleted prior to initial criticality.
e The lack of this filte~ being able to remove 90% of the organic iodines at the time of core load will not affect the ability to reduce potential consequences of design basis accidents since there is no buildup of radio-active iodines in the core (and have no pot~ntial iodine activity in the RHR System) until after initial criticality is achieved.
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ITEM ITEMS TO BE COMPLETED PRIOR TO INITIAL CRITICALITY DESCRIPTION at present, have the capability to remove 90% of the organic iodineso SAFETY EVALUATION Therefore, the absence of organic iodine removal capability will have no adverse affect on the health and safety of the
. public
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