ML20247E053
| ML20247E053 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/16/1989 |
| From: | Wilson R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| 5000-89-1767, NUDOCS 8905260107 | |
| Download: ML20247E053 (7) | |
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i GPU Nuclear Corporation arsippany, New Je sey 07054 201-316-7000 f
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a Writer's Direct Dial Nurnber:
1 May 16, 1989 q
5000-89-1767 1
Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mail Station P1-137 Washington, DC 20555 Gentlemen:
SLbject: Oyster Creek Nuclear Generating Station Docket'No. 50-219 Control Room Habitability During-the 12R outage, GPUN installed a modification that incorporates the Long Term / Final Design Objectives for control room habitability.
For reference, these design objectives from our June 4,1985, letter were as follows:
1.
The Licensee will perform a Single Failure' Analysis of the Control Room HVAC System, address all potential problem areas, and provide remedial measures. Any modifications will be implemented before the end of the Cycle 12 refueling outage and will be consistent with the criteria defined by the NRC staff at the March 19, 1985 meeting.
2.
The Licensee will assess existing diesel g; wrator capability in order to provide back up power to the final Control Room HVAC system design.
I 3.
The Licensee will meet the Beta skin dose limits with the final Control Room HVAC system design without protective clothing and goggles.
4.
The Licensee does not have to meet the natural phenomena criteria.
Previously, GPUN deferred a full response to NRC's Request for Additional Information dated February 22, 1984, to allow time for the development of this modification. At this point in time, GPUN has completed the modification so that these questions can now be addressed. We have presented the original questions with our responses in attachment 1.
By submittal of this rotponse, GPUN has complied with the Request for Additional Information, and has addressed the design objectives identified in our June 4, 1985, letter.
8905260107 890516 3
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L GPU Nuclear Corporation is a subsidiary of General Public Utikties Corporation
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. If you have any questions on this ' letter, please contact M. W. Laggart, Manager, BWR. Licensing at (201)316-7968.
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Ve y tr ly yours, 1
\\k R. F. Wil on Vice President l
Technical' Functions RFW/pa cc:
Mr.. William T. Russell, Administrator
. Region I U.S. Nuclear. Regulatory Commission -
475 Allendale Road King of Prussia, PA.
19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N.J.
08731 Mr. Alex Dromerick U.S. Nuclear Regulatory Commission Mail Station NRC Question.137-Washington, DC 20555 6374f
ATTACHMENT I RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION DATED FEBRUARY 22, 1984 NRC Question 1 Justification for all active components of the proposed control room emergency ventilation system (such as fans, isolation dampers, chillers, and radiation monitors) that will not be redundant and/or single failure proof following modification.
Response to Question 1 This request (2/22/84) for additional information preceded the meeting of March 19, 1985, (NRC Ltr. dated 4/16/85) where the need for a separate control room emergency ventilation system was negated and design objectives were established to upgrade the original system for the resolution of the Control Room Habitability issue at Oyster Creek.
Specifically, remedial measures were provided as a means to address potential problem areas which would be identified by a single failure analysis of the active components in the original control room HVAC system. These remedial measures are further discussed in the Responses to Question 4 and 5.
Since an emergency control room HVAC system meeting the single failure criterion was not our original design basis for the plant, nor was this a commitment from the March 19, 1985 meeting, the 12R modification upgrades the existing system to address the commitments in our June 4, 1985, letter. The Control Room has two independent HVAC systems for cycle 12.
Each system has four manual operating modes designated as normal, purge, partial and full recirculation. Duct mounted radiation monitors, HEpA filters, or charcoal adsorbers are nct part of either system's design.
Both systems share a common supply / return duct inside the Control Room HVAC boundary. The existing HVAC system is designated System A, and the 12R modification which includes a separate fan, dampers, and a refrigeration unit is designated System B.
System A consists of a commercial grade air conditioning unit with a supply fan, steam coils for heating, air operated dampers, and a three stage l
refrigeration unit for cooling. The auxiliary boiler supplies steam for heating, and the Turbine Building Closed Cooling Water System provides cooling water for the air conditioning unit condenser.
Diesel Generator 1 (DG-1) provides backup power to the supply fan via unit substation (USS) 1A2.
System B utilizes a commercial grade roof-top air conditioning unit with a supply f an, self-contained refrigeration unit with air cooled condenser, a duct-mounted electrical heater, and motor-operated dampers.
DG-E provides backup power to the supply fan via USS 1B3.
I NRC Question 2 l
Identify proposed technical specification requirements for periodic leak testing of the control room emergency zone, filter bypass dampers, outside air dampers and/or valves, and ESF leakage outside containment or provide justification if these leak tests are not proposed.
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~ Response to Question 2 In accordance with the criteria defined in the March 19, 1985 meeting, Technical Specification requirements for the periodic leak testing of the controi room envelope were proposed and accepted as Amendment 115, dated March 31, 1987.
A revision to the TS will be proposed to include the surveillance requirements for System B.
This proposed change will be made by a separate submittal pursuant to 10CFR50.90.
NRC Question 3 Identify the locations of radiation release points from design basis accidents other than the LOCA relative to control room outside air intake locations. To support your assessment, provide your bases and relevant layout drawings.
Response to Question 3 A response was given in Attachment III to our letter dated June 4, 1985.
NRC Question 4 Information which shows that the control room emergency ventilation system is designed to function properly in the event of a loss of offsite power or pipe breaks in areas adjacent to the control room.
Response to Question 4 As stated in our response to Question 1, Oyster Creek has two independent HVAC systems.
For a loss of offsite power (LOOP), the operating HVAC system (A or B) will trip automatically.
Since these HVAC loads are not part of the automatic loading sequence for the diesel generators, partial (supply fan enly) or full (supply fan plus air conditioning compressors) system restoration will be delayed to allow for plant stabilization and de-energization of the unnecessary safety related loads.
For the design basis event (LOOP plus LOCA with a loss of one DG), the manual loading for partial restoration (fan only) of a HVAC system is estimated to be accomplished within 30 minutes.
Full system restoration with offsite power is estimated within two hours.
Based on the review of the radiological analysis (see Response to Question 5) and the control room heat load calculations (GPUN Lir. 5000-88-1685, dated 12/12/88), the control room envelope will remain acceptable for three hours before one system.must be restored to limit the raaximum temperature to 104 F.
During the development of the 12R modification, two potential pipe breaks were j
identified in an area adjacent to the control room. The Turbine Building l
Closed Cooling Water System and the auxiliary boiler steam lines i'or System A i
are located in the Upper Cable Spreading Room. The System B refrigeratiori unit l
with outside air exhaust and recirculation air dampers, are located on the Upper Cable Spreading Rocm roof tway from these potential ?ipe breaks.
Conduits which are installed in the Upper Cable Spreading Room are routed to minimize the effects of these postulated breaks.
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p NRC Question 5 Analysis of control room operator doses following postulated design basis accidents.
It may be necessary to assess DBAs other than a LOCA if radiation release points are closer to the control room air intake than that for a LOCA.
The analysis should include a detailed listing of data and assumptions used, as well as the results.
Response to Question 5 The logic for selecting appropriate DBAs was stated in Attachment III to our letter dated June 4, 1985.
In accordance with our commitments, the assumptions and results for the radiological analysis were submitted by letters June 17, 1985 and September 29, 1986. The NRC staff accepted these assumptions for the radiological analysis by Amendment 115, dated March 31,-1987.
Since March 1987, two items prompted separate reviews of the radiological analysis for control room habitability.
First, the radiological analysis assumed the partial recirculation mode of operation for a design basis LOCA and excluded the effects of a LOOP concurrent with a DBA pending the implementation of the Long Term Design Objectives.
Second, the main steam isolation valve (MSIV) leakage rate was determined by 12R tests to vary from the leakage rate assumption used in the radiological analysis. The leakage rate is a function of the MSIV air accumulator pressure and the post LOCA containment pressure.
The radiological analysis (Ltrs. 6/17/85,9/29/86) assumed the Technical Specification limit for MSIV leakage post LOCA with leakage decreasing as a function of the post LOCA containment pressure. MSIV air accumulator pressures were not accounted for in that analysis. Therefore, the assumed behavior of the MSIV leakage rate differs from the actual behavior observed during the 12R testing.
For the first item, the radiological analysis assumed the partial recirculation mode of operation for a design basis LOCA and excluded the effects of a LOOP concurrent with a LOCA pending the implementation of the Long Term / Final Design Objectives. Since the System (A or B) trips with a LOOP concurrent with a LOCA, a review of'the radiological analysis was performed to account for partial restoration (supply fan only) of a system with 100% outside air to limit the control room maximum temperature (104 F) when ambient temperatures
-(82 F) permit this mode of operation.
The revised calculation assumed the most conservative operation (supply fan only for 30 days). Since GPUN estimates that offsite power and full System (A or B) capability will be restored within two hours, this calculation is conservative. Also, the coping analysis for. Station Blackout (GPUN Ltr.
4/17/89) shows an alternate AC power source within one hour by the end of refueling outage 14R. For the supply of 100% outside air to the contrcl room envelope, the calculated doses increase to 29.1 Rem beta and 3.14 Rem gamma for i
the assumed 30 days; however, these values are still within the allowable limits of 30 Rem beta and 5 Rem gamma.
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. For the second item, the MSIV bypass leakage is a function of the air accumulator pressure and the containment pressure post LOCA.
GPUN Letter 5000-89-1733 (3/10/89) provides the details of this MSIV leakage assessment.
The radiological analysis (Ltrs. 6/17/85,9/29/86) assumed the containment pressure would decrease to O psig in 10 days post LOCA. This assumption equated to 1000 standard cubic feet (SCF) of MSIV bypass leakage for the dose assessment.
The MSIV leakage was reassessed considering MSIV leakage as a function of
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accumulator and containment pressures. GPUN calculated a MSIV bypass leakage i
of 243 SCF for the first day post LOCA. This calculation assumed that the Drywell pressure decays to 1 psig in about 2 1/2 hours then remains constant until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the LOCA.
Since the Updated FSAR (Figure 6.2-3) shows O psig within 6 1/2 hours, this pressure profile was considered conservative for the assessment.
The MSII bypass leakage calculated by the radiological analysis (1000 SCF)
I exceeds the bypass leakage calculated by the MSIV leakage assessment (243 SCF).
S.nce the MSIV bypass leakage is the major contributor to the control room dose, the assumptions used for the radiological analysis are conservative when compared to the expected MSIV behavior (12R tests) and containment pressure response (FSAR Fig. 6.2-3) during a LOCA.
Further, the assumptions used for the main steam line volume (exclusion of main steam line header and piping volume up to the turbine stop valve) and for calculating the beta skin dose (inclusion of iodine daughter products) are conservative. Therefore, the radiological analysis of record is still bounding.
NRC Question 6 Analysis which shows that the rate of increase of toxic gas concentration in the control room will be slow enough to allow the control room occupants sufficient time to aon respiratory equipment.
Response to Question 6 Chlorine transport analyses for a one ton liquid chlorine tank cylinder were submitted with letters dated August 16, 1985 and September 29, 1986. With the removal of the one ton cylinders during early 1987, the probability of an onsite toxic gas release affecting control room habitability was minimized for Oyster Creek.
A review of the chlorine trangort analysis was performed at the new location for an instantaneous and a continuous release (3/8 inch line break) for a 150lb cylinder which 'Is stored approximately 380 feet from the control room intake, i
The analysis took no credit for mixing of the chlorine plume due to the effects i
of a building wake, and the analysis also assumed the wind direction is such that the centerline of the chlorine plume at ground level blows directly towards the control room air intake (approximately 41 foot above plant Grado).
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.. For the instantaneous and continuous releases under various meterological conditions, a toxic limit (0.0459/m ) was only achieved in the control room envelope when the control room received air from outside sources at rates greater than 13,000 cfm (instantaneous release) and 1750 cfm (continuous release until cylinder is der: ted) respectively. These control room outside air rates were assumed const.
for the duration of the accidental chlorine releases. The minimum times to achieve a toxic limit were calculated as 320 seconds (instantaneous release) and 372 seconds (continuous release). Allowing 5 seconds for the detector loop response time, the minimum operator response times to restrict the control room outside air source rate (switch to full recirculation) were assumed as 315 seconds (instantaneous release) and 367 l
seconds (continuous release).
The 12R control room HVAC tests have demonstrated an outside air source rate less than 1750 cfm for the full recirculation mode of operation (FN-826-009 off).
Further, the minimum times assumed for operator response are greater than the minimum operator response time (120 seconds) required by Regulatory Guide 1.78.
Due to these facts and the conservative assumptions noted above, we have concluded that the control room operator has sufficient time from the receipt of a chlorine alarm to place the control room HVAC system (A or B) into full recirculation, and the need to don respiratory equipment is not warranted.
I Plant procedures currently in place direct the operator to switch the system to full recirculation upon the receipt of a chlorine alarm in the control room.
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