ML20207J545
| ML20207J545 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 09/20/1988 |
| From: | Hairston W GEORGIA POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| HL-79, TAC-68893, NUDOCS 8809280032 | |
| Download: ML20207J545 (6) | |
Text
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Georo a FW*r Compary 333 P+dmont Avenue
, Atanta Georg*a 30%8 3
Te'ephcee 404 5266526 Ma+ng Address Fest Off.ce Box 4545 Anta. Georg'a 30302 W He ton, I tre sots?nn Actic sytem mcm, opmtes HL-79 0467I X7GJ17-H600 September 20, 1988 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Hashingtor,, D.C.
20555 PLANT HATCH - UNIT 1 NRC DOCKET 50-321 OPERATING LICENSE DPR-57 REW ESI TO REVISE TECHNICAL SPECIFICATIONS:
SECONDARY CONTAINMENT BOUNDARY RIQUIREMENTS DOCUMENTATION OF TELEPHONE CONVERSATIONS l
Gentlemen:
l On July 11, 1988, Georgia Power Company (GPC) submitted a request to l
the NRC to revise the Plant Hatch Unit 1 Technical Specifications relative to secondary containment boundary requirements.
This letter documents telephone conversations, related to that submittal, held from August 24, through September 13, 1988, between NRC reviewer Charles Tinkler, Southern Company Services, and GPC personnel.
The answers provided summarize our conversations, which we understand resolve the NRC concerns relative to the subject submittal.
NRC.00ESTION:
Why is doing the HSIV work causing a redefinition to be required?
l Is there other work which necessitates the change?
I GPC RESPONSE:
Outage plans may require simultaneous work on the main steam isolation valves (MSIVs) and turbine stop/ control valves. The removal cf i
components on the rain steam lines (e.g.,
valve parts) could possibly
.,og make pulling a
reactor building vacuum difficult or impossible, ab Additionally, work is planned on the high pressure coolant injection o,
(HPCI) and feedwater drain lines which could also create a leakage path gh from the reactor building.
The requested change, then, is driven by the need to optimize outage planning and reduce outage costs.
g l
ka:
NRC_WESILOB:
4 aor What is the historical basis of the secondary containment boundaries I
$$a.
during refueling; 1.e.,
why was the Reactor Building of the unit refueling even included?
Is it because the normal configuration allows Acol for casy access?
If so, why isn't the access necessary now?
4 I
t
Georgia Power d U.S. Nuclear Regulatory Commission September 20, 1988 Page Two GPC RESPONSE:
Under the normal Unit I secondary containment configuration, which includes the common Unit I and Unit 2 area above the refueling floor and the Unit I reactor building area below the refueling floor, items can be moved up through the open hatch to the refueling floor.
(Reference Figure 1 included in GPC's July 11, 1988, submittal.) This configuration would be used in the first part of the outage.
During the latter part of the outage, the modified configuration, which includes only the common Unit I and Unit 2 area above the refueling floor, would be used.
(Reference Figure 2 included in GPC's July 11, 1988, submittal.)
Naturally, secondary containment integrity would be demonstrated after changing to the modified secondary containment configuration.
(Reference proposed Technical Specification 4.7.C.2 included in GPC's July 11, 1988 submittal.)
NRC_OUESIl03:
Reference Figure 2 (Sheet 2 of 2) included in GPC's July 11, 1988, submittal:
1.
What type of valves are F032A & B?
2.
What type of valves are F334A & B?
3.
What it meant by "closed and tagged out" for the different valves?
4.
What are the single-failure considerations for the isolated lines?
GP_C RESPONSE:
1.
Valves F032A & B are air-operated 18-in, damper; (normally closed-fail open).
2.
Valves F334A & B are 2-in, air-operated globe valves for the Centainment Atmosphere Dilution (CAD) system.
These valves are normally closed (NC) and fall closed (FC).
3.
Administrative procedures will specify how the valves are "closed and tagged out." Generally, tags would be placed on the valve itself and at locations where the valve could be activated.
If the valves are powered and can be remotely actuated
- ontrol switches are placed in the closed position and their solenoids depowered.
Dampers F032A & B, which fail open, 04671
Georgia Power d U.S. Nuclear Regulatory Commission September 20, 1988 Page Three GP_C_BESPONSE: (Continued) will be physically blocked closed.
Valves F312 & 313 which are manual valves will be closed and tagged out.
The remaining valves, F332 A&B, F334 A&B, F338 and F340, are valves which fail closed.
The control switches for these valves will be placed in the closed position, their solenoids will be depowered, and each valve will be tagged out at locations where the valve could be remotely activated (e.g.,
local panel or in the main control room).
4.
Based on the administrative controls and physical constraints that will be imposed to secure these valves closed for the isolated lines, single-failure considerations do not apply, and the ability of the standby gas treatment system (SGTS) to maintain secondary containment integrity will be verified per Technical Specifications (TS) requirements.
MC_0UISILON:
Hill both trains of SGTS be operating during testing?
Do the surveillance requirements per 3.7.C.I.a and 3.7.C.2.a mean that rech train is tested to pull 1/4-in, vacuum, or both together?
Is the SGTS single-failure groof?
Why does the Unit 1 FSAR state the SGTS design f)3w as 3000 f ta per minute and the TS surveillance requiretrent refers to 4000 ft3 per minute.
GPC RESPONSE:
As discussed in Unit 1 TS bases 3.7.8 (pages 3.7-34 and 3.7-34a) operation of one of the two Unit 1 SBTS trains and one of the two Unit 2 SGTS trains are required to achieve the design differential pressure of 1/4-in. Accordingly, when testing is performed per TS requirements, Unit 1 SGTS train A and Unit 2 SGTS train A are tested together.
- Likewise, both units train B SGTS are tested together to demonstrate compliance with TS requirements.
The system is single-failure proof.
Regarding the Unit 1
SGTS flow design, the Unit 1 FSAR is misleading; however, GPC plans to revise the wording in the next FSAR annual update.
The 3000 f t3 per minute was a nominal operating point for the Unit I fans, but with proper damper adjustment, the fans are capable of flows in excess of 4000 ft3 per minute.
However, the TS limit of 4000 f t3 per minute ensures the required filter efficiency is maintained and tests the adequacy of the secondary containment integrity.
04671
J Georgia Power d U.S. Nuclear Regulatory Commission September 20, 1988 Page Four NRC OUESTIOf(:
The analysis in the Hatch Unit 2 FSAR, section 15.1.41 used a volume of 2,840,000 ft3 for the effective volume in the reactor building following a fuel handling accident.
This volume resulted in an air change rate of 3.6 air changes per day at the SGTS rated flow of 7000 ft3 per minute.
The prpposed Technical Specifications change would remove the 1,275,000 ft3 of reactor building volume below the refueling floor from the effective volume in the analytical model.
What will be the impact to the fuel handling accident realistic dose analysis presented in Hatch 2 FSAR section 15.1.41 relative to this reduction in the effective building volume?
GPC RESPONSE:
[
A review of Hatch 2 FSAR section 15.1.41 indicates the following:
A.
This section contains two analyses, one based on realistic conservative engineering assumptions and the other on more conservative NRC assumptions included in Regulatory Guide 1.25, dated March 1972.
A comparison of these analytical assumptions is outlined in FSAR Table 15.1-48.
The resulting dose effects e
from these analyses are as follows:
Realistic (Conservative) Engineering Assumptions (Reference FSAR Table 15.1-45.)
f Whole-Body Dose (rem)
Inhalation Dose (rem)
Exclusion area 7.7 E-06 2.9E-04 LPZ 4.3 E-05 3.1E-03 NRC Conservative Assumptions (Reference FSAR Table 15.1-47.)
Whole-Body Dose frem)
Inhalation Dose (rem)
Exclusion area 3.7 E-03 0.256 i
LPZ 4.3 E-03 0.264 B.
Both fuel handling accident analyses are based on 3.6 air changes per day or 360-percent building leakage. 2,840,000 f t}
If the fue handling area effective volume 1:; reduced from i
3 to 1.565,000 f t, the 3.6 air changes per day will increase to 6.4 air changes per day, increasing the resulting dose by a 0467I
Georga Power d U.S. Nuclear Regulatory Commission September 20, 1988 Page Five L
GPC RESPONSE: (Continued) factor 1.78.
For example, using the higher doses based on the NRC analytical methodology, the dose impact would be as follows:
Whole-Body Dose (real Inhalation Dose (rem)
Exclusion area 6.6 E-03 0.455 LPZ 7.6 E-03 0.469 Comparison of these dose increases to the 10 CFR 100 guidelines (25-rem whole-body; 300-rem thyroid dose) indicates the j
following impact percentages:
2: cent of Limit Increase i
htn a-Body Dose Inhalation Doit
[
Exclusion area 0.026%
0.15%
LPZ 0.03%
0.16%
l C.
Additionally a review of the current fuel handling accident requirements outlined in the Standard Review Plan (NUREG-0800),
Section 15.7.4, Item II.1 indicate dose mitigating engineered safety feature systems are acceptable with respect to radiological consequences if calculated whole-body and thyroid doses at the exclusion area and LPZ are well within the exposure guidelines of 10 CFR 100.
"Hell within" is defined as 25 percent or less of the 10 CFR 100 exposure guideline values.
Furthermore, Item II.3 indicates using conservative assumptions I
in Regulatory Guide 1.25 is an acceptable methodology of calculating the dose effects of a fuel handling accident.
i Based on the evaluation presented above, where the highest dose impact is 0.16 percent of the 10 CFR 100 guideline dose values and the i
l limiting acceptable value is 25 percent or less of the guideline values, i
the radiological impact of reducing the fuel handling area effective volume can be considered insignificant.
l If you have questions regarding these responses, please contact this office at any time.
L Sincerely, j
u/. $. /[4.,
I H. G. Hairston, III Distribution:
See Page Six t
0467I
Georgia Power b U.S. Nuclear Regulatory Commission September 20, 1988 Page Six GKM/ac c: Georaia Power Company Mr. H. C. Nix, General Manager - Hatch Mr. L. T. Gucwa, Manager Licensing and Engineering - Hatch GO-NORMS U.S. Nuclear Reaulatory Commission. Washinoton. D.C.
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Mr. L. P. Crocker, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaton II
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Dr. J. N. Grace, Regional Administrator Mr. J. E. Menning, Senior Resident Inspector - Hatch i
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