ML20045E692
| ML20045E692 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/29/1993 |
| From: | Borchardt R Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9307020326 | |
| Download: ML20045E692 (8) | |
Text
Q f
+
une 29, B93 Docket ~No.52-001 V
. Mr.
ick W. Marriott, Manager Licen..ng &. Consulting Services GE Nuclear Energy
'175 Curtner Avenue San Jose, California 95125
Dear Mr. Marriott:
SUBJECT:
COMMENTS ON INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA (ITAAC) AND RESPONSES TO LETTERS DATED JUNE 4 AND 22, 1993 is 'a list of comments regarding the staff's review of ITAAC.
GE-should revise the ITAAC to resolve these comments, and provide a mark-up immediately. Timely responses and limited iterations are essential to meeting the schedule.
You will receive additional comments on ITAAC submittals in the-near future.
Enclosures 2 and 3 are the staff's responses to GE's letters dated June 4 and 22, 1993, respectively.
Eaclosure 2 concerns insights from the ABWR severe accident analysis, while Enclosure 3 is on the containment overpressure protection system.
Sincerely, (Original signed by)
R. W. Borchardt, Acting Director Standardization Project Directorate 1
Associate Directorate for Advanced Reactors and License Renewal d
Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures:
See next page DISTRIBUTION:
'~ Docket File PDST R/F TMurley/FMiraglia
.DCrutchfield NRC PDR PShea JNWilson SKoenick SNinh CPoslusny-RBorchardt TBoyce-
,DTang RJones, 8E23 AThadani, 8E2 HRichings, 8E23
' ACRS (11)(w/o enc)JMoore,15B18 GGrant, 17G21 BHardin, RES LShao, RES AVietti-Cook WRussell, '129)8 J0'Brien, RES 0FC-LA:PDST PE:PDST PM:Pp}T
'SCPN (A)hDST NAME- 'PShea /k bSKoenick:Y CPo$ny bhiIson RBkhardt b/ /h3 h/2V93'
/26/93-(# fd/93 l0/h93--
N DATE
({-
OFFICIAL RECORD COPY:
LGITC628.SK.
t 9307020326 930629
.PDR ADOCK 05200001 A
'PDR.
1
~
Mr. Patrick W. Marriott Docket No.57-001 General Electric Company cc:
Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear Energy
.175 Curtner Avenue General Electric Company
~I San Jose,. California 95125 175 Curtner. Avenue, Mail Code 782=
San Jose, California 95125 Mr. L. Gifford, Program Manager Regulatory Programs er unclour Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U.S. Environmental Protection Agency 401 M Street, S.W.
Washington, D.C.
20460 Mr. Sterling Franks U.S. Department of Energy NE-42 Washington,.D.C.
20585 Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004' Jay M. Gutierrez, Esq.
Newman & Holtzinger, P.C.
1615 L Street, N.W.
Suite 1000 Washington, D.C.
20036 1
<.1 2.8.1 Fuel Bundle Desian Description 1.
The first paragraph should be removed and replaced with a description of the system purpose of the fuel'. This should contain.a definition of what constitutes a fuel bundle, and the relationship to the fuel channel.
2.
Remove from the second paragraph " supplied to any facility utilizing ~ the certified design" and -add "and is evaluated using methods _ and criteria:to assure that:" Remove the third paragraph.
3.
Item (1) may be interpreted to indicate that if a fuel failure occurs for any reason during normal operation, the fuel design is not in compliance with this criteria.
Improvement of the language for clarification should-be considered.
4; Add to Item (4) "for all design bases events, including seismic and LOCA events."
5.
The standard safety analysis report (SSAR) needs to be amended to support this design description.
2.8.2 Fuel Channel Desian Description 1.
Add an Item (3) indicating channel bowing for the channel design lifetime will not cause specified acceptable fuel design limit to be exceeded during normal operation and anticipated operational occurrences.
2.
Indicate in the first paragraph any other channel purposes, such as control rod insertion.
3.
Address Standard Review Plan 4.2 Appendix A requirements for seismic events in Item (1).
2.8.3 Control' Rod Desian Description A.
Design Description 1.
Indicate scram reactivity insertion for safety shutdown response as a function to be considered.
B.
Design Cri_teria 1.
Seismic requirements should be added.
2.
Indicate the need to consider two stock control rods because of the rod accumulator design.
3.
Design criterion'(4') should be replaced with functional criteria which embody the general' design criteria (GDC) requirements such as those
-indicated in GDC 25 through 29.
Enclosure ' l
2.8.4 Loose Parts Monitorina System (LPMS) Desion' Description
-1.
The design description should indicate the requirements for operability following seismic conditions as provided in Regulatory Guide 1.133.
ITAAC 1.
The ITt. fer Design Commitment Number 2 states inspection, yet the acceptance criteria clearly can only be demonstrated by test.
Please modify the ITA column to include the appropriate test description, and assure the test is provided in Section 14 of the SSAR.
2.
Add the control room configuration ITAAC for this system.
n
J Response to GE letter dated June 4, 1993
Subject:
Insights from the ABWR' Severe Accident Analysis Tier 2 - Containment Overpressure Protection System The following sections of SSAR, relating to the COPS, should remain intact:
6.2.5.1(17)-
6.2.5.2.1 6.2.5.2.6 6.2.5.3 6.2.5.4 Table 6.2-7 19E.2.3.5 19E.2.8.1 19K ll.6 Tier 1 - Containment Overpressure Protection Systems Design Description - The containment overpressure protection system (COPS) is-part of the atmospheric control system and provides containment overpressurization relief following severe accidents.
The system removes steam and non-condensible gases from the wetwell airspace to the plant stack and has a capacity of removing steam equivalent to at _least 2% rated reactor power. The COPS consists of piping from the wetwell airspace, containment isolation valves, and a rupture disk. The system is used prior to containment pressure reaching ASME Service Level C limits and can be isolated following actuation.
ITAAC Design Commitment:
1.
The basic configuration of the COPS is as shown on Figure xx.xx.
2.
The COPS has a capacity to relieve steam flow, at the opening rated value, corresponding to greater than or equal to 2% rated reactor power.
3.
The twe containment isolation valves in the COPS fail open on loss of pneumatic pressure or loss of electrical power to'the valve actuating solenoid. Any position other than full open alarms in the control-room.
4.
The COPS rupture disk will actuate at a pressure above containment design pressure and below ASME Service Level C limits.
5.
The COPS vent path up to the plant stack is designed to withstand pressures associated with system operation and is capable of preventing'or accommodating a hydrogen detonation in the pathway..
~
6.
The. containment isolation valves are. capable of fully opening and closing at pressures up to the COPS actuation pressure.
Encl osure -
c
, Inspection, Tests, Analyses:
1.
Inspections of the as-built system will be conducted.
2.
Inspection and a relief capacity calculation using as-built dimensions and loss coefficients will be performed.
~
3.
Tests will be conducted on the as-built containment' isolation valves.
4.
Inspect the manufacturer documentation of the rupture disks.
5.
Inspect the provisions or analysis which ' demonstrate' the ability of the COPS to prevent or accommodate a detonation.
6.
Inspect manufacturer documentation of the containment isolation valves.
Acceptance Criteria:
1.
The as-built COPS conforms with the basic configuration'shown in figure xx.xx, 2.
The relief capacity of the COPS is greater than or equal to 2% rated reactor-power.
3.
The containment isolation valves fail open on loss of pneumatic i
pressure or loss of electrical power to the valve actuating solenoid.
Alarms in the main control room indicate any position other than full open.
4.
The rupture disk have been manufactured to open atLthe rated burst pressure and the necessary tests have been performed certifying this.
~
~
5.
The COPS design has prevented a detonation from occurring or can handle the loadings associated with a detonation.
6.
The containment isolation valves are designed to fully open and close at the COPS actuation pressure.
b f
Response to GE Letter dated June 22, 1993
Subject:
Submittal ' Supporting Accelerated ABWR Schedule - Containment Overpressure Protection System Addresses: Open Items 20, 27, and 31 of June 7-10, 1993, meeting Comments on Open Items 27 and 31 (Containment Overpressure Protection System) 1.
Has the vent line size been changed from 14-inches (p. 12) to 250 mm (p.19)? NRC acceptance was previously based on the line sizing of 14--
inches capable or relieving steam equivalent to-2.4% rated power.
Provide justification for any change in line size.
2.
Explain the rationale for specifying ac power to the containment isolation valves (F007 & F010) of the COPS, as opposed to de power. The Commission accepted a rupture disk on the basis that control over the venting process exists.
In a severe accident, the potential for availability of ac power is less than that of de power.
3.
What is the purpose of valve F0597 4.
On Table 3.9-8, why are valves F007 and F010 not indicated as isolation-valves and why is valve F007 not stroke tested?
5.
Valves F001 through F0ll are all containment isolation valves. Why is the test frequency different for some valves (i.e., 2 yrs,' 2. yrs 3 mo,- and 2 yrs R0)?
6.
Do you need to change Table 3.9-8 for valves D001 and D0027 Comments on Open Item 20 (Vent and Purge Lines) 1.
Provide the basis for not providing separate containment penetrations _for lines which penetrate the containment (i.e., 50 mm nitrogen supply line e
and 550 mm. purge line).
2.
Provide the basis for not putting one containment isolation valve inside containment and one outside containment.
3.
Provide the basis for selection of the valve closure time of less than
'20 seconds for the purge and vent valves. The basis for SRP-selection of' a 5-second closure time was to decrease the time to as low as practical.
What is the shortest time practical for these valves to close based on industry experience?
4.
What type of barriers' are used to prevent debris from entering the purge and vent' lines'and affecting the valve seating characteristics? What was the sizing criteria for the barriers and what criteria is used in selection of the piping within containment to which the barrier'is attached?
4
fi 4 5 Do the fechnical Specification provisions fo'r the vent and purge lines..
agree with this write up? Where-is tha criteria for less than 15% power?
6.
Although the likelihood of a LOCA during inerting/de-inerting may be low, the potential for a plant transient during these type of evolutions (coming up or going down in power) are not.
/. Discuss specifically which valves will'be used for feed and bleed operation and which ones will be used for vent and purge.
What controls are placed over operation of the large valves?
I 8.
What number of supply and exhaust lines are used at any one time?
9.
Do any of the valves have resilient seats?
- 10. What provisions are provided for reverse testing of the purge and vent valves?
9 8
r i