ML20247N325
| ML20247N325 | |
| Person / Time | |
|---|---|
| Issue date: | 09/14/1989 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | Hall W NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| NUDOCS 8909260191 | |
| Download: ML20247N325 (9) | |
Text
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- = . Mr.:Farren J. Hall. Manager September 14, 1989
' Opet 4cns, Management, and
- . . Suppe : Services Division.
Nuclear Management and Resources Council 1776 Eye Street, Suite 300 Washington, D.C. 20006-2496
Dear Mr. Hall:
SUBJECT:
NRC STAFF REVIEW 0F THE OWNERS GROUPS' PROPOSED NEW..
STANDARD TECHNICAL SPECIFICATIONS (STS), SECTION 2.0 Enclosed is the marked-up Safety Limit Section (2.0) for the BWR Owners Group submittal and associated NRC staff comments. The NRC staff finds the GE format for section 2.0 preferable. The WOG, CE03, and B&WOG should convert the LC0 format for Section 2.0, Safety Limits, to a paragraph format (similar to the BWR submittal). This is a perceptual cue to the operator that Safety Limit violations must be addressed differently than LCO violations. For example, throughout Section 3 of the improved STS, a condition is typically no longer applicable if all the associated Required Actions are complied with. Thus, a new Condition would be entered (with new Required Actions) or the unit would be back in compliance with the LCO. In the case of a Safety Limit violation, the Condition is not left even if the Safety Limits are restored. Restart of the unit is not allowed until authorized by the Commission. This will avoid confusion and potential misinterpretation resulting from the differences in noncompliance between LCOs and Safety Limits.
Enclosed are general comments related to the new STS submittals. Also enclosed :
are specific comments related to 1.he new STS submittals.
In accordance with our review plan, the NRC staff is prepared to discuss its comments marked on the proposed new STS, jointly with the four Owners Groups. ;
The Owners Groups are requested to coordinate with the staff,'through NUMARC, a ;
time for the meeting.
OIt'dHETGilfs00SE& CALVO Jose A. Calvo, Chief 1 Technical Specifications Branch Division of Operational Events Assessment Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission i
Enclosures:
NRC Staff General Comments Related to the New STS Submittals (PWR's) )
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o Ms. J. R. Hinds.
Westinghouse Owners Group.
'c/o Pacific Gas & Electric Co.
Diablo Canyon Power Plant
'P.O. Box 56-Avila Beach, CA 93424 Mail Stop 104/5/21B-
.Mr. R..A. Bernier CE Owners Group c/o Arizona Public Service 11226 North 23rd Street
. Phoenix, AZ 86072 Mail Stop 7048 Mr. C. W. Smyth Babcock & Wilcox Owners Group c/o GPU Nuclear Co.
TMI Nuclear. Station P.O. Box 480 Middletown, PA 17057 Mr. R. R. Sgarro GE~(BWR) Owners Group c/o Pennsylvania Power & Light 2 North Ninth Street Annex 2-4 1, Allentown, PA 18101 l
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Enclosure l
NRC Staff General Comments Related to the New STS Submittals (1): For PWR's it is acceptable for the Reactor Core " Safety Limits" section to consist of the actual Fuel Centerline Temperature Safety Limit, DNBR Safety Limit, and other applicable Safety Limits. The curvas associated with the Reactor Core Safety Limits (i.e. W figure 2.1.1-1 Reactor Core Safety Limits, CE-analog figure 2.1.1A-1 Reactor Core Thermal Margin Safety Limit, B&WOG figure 2.2.1-1 Reactor Core Safety Limit) which specify operat-ing restrictions, imposed so that the actual Safety Limits are not exceeded, can be relocated to the COLR provided they are appropriately referenced by
.an LCO. The LCO may also require an outlet quality parameter or hotleg boiling. limit.
For example: Section 2.0 states actual Safety Limit.
Section 3.0 requires compliance with operating curves (that prevent the limit from being exceeded)
COLR (or other programmatic document which is subject to NRC staff review, though not prior approval, when changed for each Cycle) provides curves.
(2) The Bases have not yet been completely reviewed,- and therefore, specific comments are not provided. However, the following are general comments related to the Bases.
- a. The Bases should reference the applicable FSAR sections.
[' The NRC Staff prefers the way B&WOG wrote the " Applicability" section.
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NRC Staff Specific Comments Related to the New STS Submittals (PWR's)
CE0G Submittal (1) The curves associated with Safety Limit "2.1.1 Reactor Core" (Figure 2.1.1A-1 which can be relocated to the COLR provided it is referenced by an LCO) should address flow requirements (operating pumps / loops).
(2) At the ent of Safety Limit 2.1.1 DNBR add, "as determined by the [ ]
correlatica."
(3) Bases, p. B 2.2-1 under " Applicable Safety Analysis", states that there are no accident analyses that assume the RCS reaches its pressure safety limit during any design bases event. The NRC Staff prefers the attached Westinghouse write-up.
(4) The Bases for the Peak Linear Heat Rate should address burn-up dependencies and the specific number provided in Safety Limit 2.1.2.
B&WOG Submittal In the Bases, p. B 2.2-2 under " Applicable Safety Analysis", the last paragrsph relates to " Axial Power Imbalance." It should be relocated to the Bases for Axial Power Imbalance Safety Limits, or be deleted.
WOG Submittal The curves associated with Safety Limit 2.1.1 " Reactor Core Safety Limits" (Figure 2.1.1-1 which can be relocated to the COLR provided it is referenced by an LCO) should address flow requirements (operating pumps / loops).
1 NRC Staff Comments to BWR New STS Submittal Section 2.0 Safety Limits (Note Mark-up)
(1) . In the Bases, address the applicable conditions to safety limits.
(2) To convey the necessary urgency, add the word " Initiate the following action immediately, and complete", to the beginning of safety limit 2.2.2.
(3) Change safety limit 2.2.5 to read:
"The reactor mode switch shall be placed in only the Shutdown or Refueling positions. Restart of the Unit shall not commence until authorized by the Commission.
(4) In the Bases description on page B 2.0-1 under " Fuel Cladding Integrity":
in the first sentence replace the word "or" (between "785 psig" and " core flows") with the word "and" ' and, in the second sentence replace the word "and" (between " low pressures" and " low flows") with the word "or". These changes will make the Bases description consistent with the safety limits
~in 2.1.1 and 2.1.2.
(5) . The reference section of the Bases on page B 2.0-5 should include appropriate explicit references to the FSAR.
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Safety Limits )
2.0- 1
, 2.0 SAFETY LIMITS
[ 2.1 SAFETY LIMITS l l
2.1.1 With the reactor steam dome pressure < 785 psig or core flow
< 10% cf rated core flow:
THERMAL POWER shall be 125% of RATED THERMAL POWER.
2.1.2 With the reactor steam dome pressure 1785 psig and core flow 110% of rated core flow:
MINIMUM CRITICAL POWER RATIO shall be 1 [ 1.07 ) for two loop recirculation or 1 [ 1.08 ) for single loop recirculation operation.
2.1.3 Reactor vessel water level shall be > the top of active irradiated fuel.
2.1.4 Reactor steam dome pressure shall be 1 1325 psig.
2.2 SAFETY LIMIT VIOLATION With any Safety Limit not met the following actions shall be met:
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2.2.1 Within one hour notify the NRC Operations Center in
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accordance with 10CFR50.72.
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2.2.2 A ithi two houri: ~
A. Restore compliance with all Safety Limits, and B. Insert all insertable control rods.
2.2.3 The [ General Manager - Nuclear Plant and Vice President -
Nuclear] and the [SRC] shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.2.4 A Licensee Event Report shall be prepared pursuant to 10CFR50.73. The Licensee Event Report shall be submitted to the Commission, the [PRB], the [ SRB] and the [ General Manager
- Nuclear Plant and Vice President - Nuclear] within 30 days of the violation.
2.2.5]d! : :;:n: of the unit shall not h^$ $ N until authorized by the Commission.
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2-1 4/28/89 u_._-____m______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
RCS Pressure Safety Limit-B 2.1.2-B 2.1 SAFETY LIMITS B 2.1.2 Reactor Coolant System (RCS) Pressure Safety Limit BASES BACKGROUND The restrictions of this Safety Limit protect the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, its continued integrity is assured.
-The design pressure of the RCS is [ ] psia. As an assurance of system integrity, all RCS components are hydrostatically tested at 125% of design, [ ] psig, per the ASME code requirements prior to initial operation when there is no fuel in the core. Should repairs or replacements take place which would require a full hydrostatic test of the RCS, the fuel would have to be completely off loaded in order to exceed the maximum pressure specified in this LCO. Without fuel in the core there is no chance of fission products getting into the reactor coolant.
Overpressurization of the RCS could result in a breach of the RCS pressure boundary. If this occurred in conjunction with a fuel cladding failure, fission products could enter l the containment atmosphere raising concerns relative to limits on radioactive releases specified in 10 CFR 100.
APPLICABLE The RCS pressurizer sa'fety valves are sized to prevent SAFETY ANALYSES system pressure from exceeding the design pressure by more than 10% in accordance with Section III of the ASME Code for Nuclear Power Plant components (Ref.1). The transient which establishes the required relief capacity, and hence valve size requirements and lift settings is a complete loss of external load without a direct reactor trip.
(continued) l l
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1 B 2.1-6 Revision l Unit Name Rev. A WOG MERITS 4
1 RCS Pressure Safety Limjt' ;
.. B 2.1.2 BASES l'
APPLICABLE During the transient, no control actions are assumed except SAFITY ANALYSES- that the safety valves on the secondary plant are assumed 3 (continued) to open when the steam pressure reaches the secondary plant safety valve settings and nominal feedwater supply is maintained.
. i More specifically, no credit is taken for operation of the following:
Pressurizer Power-Operated Relief Valves (PORVs),
Steam line relief valve, Steam Dump System, Reactor Control System, Pressurizer Level Control System, or Pressurizer spray valve. ...
The RCS pressurizer safety valves, the main steam safety valves, and the reactor high pressure trip have settings established to assure the RCS pressure safety limit will not be exceeded (Ref. 2).
SAFETY The maximum transient pressure allowable in the RCS LIMIT pressure vessel under the ASME Code,Section III is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings under [USAS Section B31.1) is 120% of design pressure. The most limiting of these two allowances is the 110% of design pressure; therefore, the safety limit on maximum allowable RCS pressure is established at [ ] psig.
APPLICABILITY Safety Limit 2.1.2 applies in MODES 1 through S because it is conceivable to approach or exceed this Safety Limit in these MODES due to overpressurization events. The Safety Limit is not applicable in MODE 6 since the reactor vessel
' head closure bolts are not fully tightened making it impossible to pressurize the RCS.
(continued)
B 2.1-7 Revision Unit Name Rev. A WDG-MERITS I _ _ _ _ _ _ - - _ _ _ _ _ _ _ -
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DISTRIBUTION Central Files OTSB R/F n - -CERossi DOEA R/F.
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~JACalvo _CJHaughney DCFischer : PDR RLEmch OTSB Members
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