ML19345D982
| ML19345D982 | |
| Person / Time | |
|---|---|
| Issue date: | 10/21/1980 |
| From: | Schroeder F Office of Nuclear Reactor Regulation |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8012220311 | |
| Download: ML19345D982 (7) | |
Text
C5 + Uw.1.-
da.-
~
00T 211980 MEMORANDUM FOR: Darrell G. Eisenhut, Director Division of Licensing FRCH:
Frank Schroeder, Acting Director Division of Safety Technology
SUBJECT:
PLANTS WITHOUT SEISMICALLY QUALIFIED AUXILIARY FEEDWATER SYSTEM
REFERENCES:
1.
Memorandum for D. G. Eisenhut fmm R. J. Mattson, dated August 8, 1980 2.
Letter for W. J. Dircks fmm M. Plesset, dated October 15, 1980 This past spring, DST, with input from other Divisions, began an analysis of the proble of plants without seismically qualified auxiliary feedwater systems.
A brief presentation was made to the ACRS on this subject on June 4, and the staff promised a briefing on the results of a preliminary safety assessment of such plants in three months. The results of DST's preliminary analysis. of the risk of core melt from such plants, and our recommendations, were provided to DL on August 8 (Reference 1). These results, and the staff's current plans for the further, more detailed, plant specific assessment of this probis, were the subjects of recent ACRS briefings held on October 8 and'10.
The first sentence of recomendation 2 of our August 8 memorandum stated, "We recomend that a more detailed study of this question be undertaken by NRR, managed by DL, and completed in the next several months." This recomendation, which clearly implied completion of the detailed assessments by around November, was discussed extensively with members of your organi::ation before Reference I was issued. The recent ACRS letter (Reference 2) also supports the judgment
...that high priority should be given to resolution of this matter."
It now appears, based on the content of the recent ACRS briefing, that the more detailed study realistically will not be completed by DL for at least another six or seven months. We believe that this time schedule does not comport with the potential safety significance of the probim. Therefore, the staff should seek a more expeditious resolution. While the DST analysis demonstrates to our satisfaction that plants do not have to be shut down while the more detailed analyses proceed, the several month timeframe was not envisioned to extend until
^
April or May of next year.
f ',,
- f. - t l' D OFFICE SURNAVE.
012-2 e o
!. 31I-DATE..
.h.
.l.
NRC FORM 318 <9 761 NRCV 324"
^
- 8 5 GOV E #N"ENT POINTING OFFICE: 1979 289 369
.a
~
-w D. G. Eisenhut Also, as a matter of clarification, the three-year timeframe discussed in our preliminary risk analysis was not intended by DST as a Loal for taking final corrective action. Such actions should be taken as quickly as reasonable; and it was envisioned that some plants would require rather prompt, interim remedial action in order to attain a reasonable level of decay heat removal reliability for seismic events. The assessment of risk over the next three yer.rs was perfomed only to give an idea as to the relative magnitude of' risk associated with this problem so that some reasoned judgment could be made as to whether prompt action must be taken, or whether (at least for some plants).
the increase in risk was small enough to await ccmpletion of longer-tem pro-grams (TMI Task Action Plans II.E.3.2, 3, and 4).
The current plan of action probably cannot be expedited much. Therefore, in view of our concern regarding the slippage of the schedule for perfomance of the more detailed studies and the ACRS concerns regarding the time for better quantification of the risk on a plant-specific basis and for the taking of any practical remedial actions, we believe that NRR should relock at the pro-posed program to see whether safety improvements could be more quickly achieved while still retaining some of the benefits of a reasonably deliberate pace.
Toward this end, we believe that there would be merit to issuing a requirement to licensees with plants not having a seismically qualified auxiliary feed-water system to either take steps to seismically qualify the auxiliary feed-water system, or to begin imediately to analyze, design, and install a seismically and environmentally qualified " bleed and feed" or " feed and bleed" system to assure the capability for decay heat removal in the event of a severe seismic event.
To expedite the installation of such a system, design and analysis criteria could be specified in the very near future, and the affected licensees instruc-ted to install an appropriate system (or upgrade the auxiliary feedwater system) without pre-implementation review by the NRC staff. We have enclosed our preliminary views on possible suitable criteria for feed and bleed systems.
In addition, realistic criteria would also have to be developed to define an interim acceptable auxiliary feedwater system. Such AFS criteria are necessary both to identify those plants that need upgrading and as acceptance criteria for the modified systems, recognizing that it would be unrealistic to upgrade old systems to meet all present requirements for new plants.
A post implementation audit review could be conducted by a NRR/I8E group to confim that the installed systen satisfies the specified criteria. Such an approach would result in a faster schedule for making needed system improve-ments and would also result in less manpower required for staff review.
Af ter you have reviewed our corsnents and suggested criteria, we request that DL arrange a meeting at the earliest convenience to discuss these matters with OST and other appropriate NRR Divisions.
gg Frank Schroeder, y l
Frank Schroeder, Acting Director Division of Safety Technology emcc h.
.f.
. ).
w%wc k.
]..
j
.p I
one E.
. j.
[
e a
.]
y i'
D. G. Eisenhut
Enclosure:
As stated cc:
H. Denton E. Case S. Hanauer D. Ross R. Vollmer i
l R. Purple T. %vak M. Ernst G. Holahan R. Baer Distribution:
Central Files SPEB Rdg.
F. Schroeder I
r A
Rt.Ba
' der' 5,;,s carc> 10/1//80 10/1A/80 10/Q)/.80
.9 N R C FO R M O ld { #- 76) N R C'.t 0 *43 C ; s, coyc r:.4.,,E"47 AR4NTING OFFICES 1979-289 369 g
i Enclosure Draf t Criteria for Bleed and Feed Recuirements 1.
Hardware The PCRV(s)1/ and all associated power supplies, controls, and posi-a.
tion indication instrumentation shall be designed to meet the folicwing j
requirements:
(1) Seismic Category I.
(2) Environmentally qualified to the most limiting' combination of tenperature, pressure, humidity and radiation expected af ter any design basis event or after several hours of use of the bleed and feed system.
(3) Operable from the control rocm.
2
\\
l (4) Powered from the onsite power supply. The overall system need not I
j satisfy the single failure requirement, however, there'shall be provisions to supply power from either emergency bus from the con-i trol room. Suitable interlocks shall be provided to prevent inter-connection of the_ emergency buses.
i b.
HPI pumps, associated valves, and all associated power supplies, i
supporting subsystems (e.g., room coolers), instrumentation and con-trols must satisfy the following requirements:
(1) Seismic Category I (2) Operable from control room.
1 IIThe term "PORV" does not preclude the use of some other type of valve to perform the bleed function.
4 1^
1 4-,--
9.=-*Fr*
-~*'"W--v"w
"~~~---*=*+-'WV-t 9~-N----
w---e t<=w--r
'*e'--r-F~vc*w'-"v-++m-*=v-+-wn-*
T-*v--*-t---we
- *-~-'
1 l.
(3) Satisfies the single failure requirement.2/
(4) Any valves inside ccntainment should be environmentally qualified to the same conditions specified in a.(2), above.
2.
Analysis The bleed and feed system is to be analyzed using a ccmputer code that satisfies i
the provisions of Appendix K to 10 CFR 50. The analysis need not consider a
~
single failure of the PORV system, except that the system must be isolatable.
The analyses required will be dependent on the characteristics of the high pressure injection (HPI) pumps (flow rate versus head and the pump shutoff 1
head), and the number and size of the FORVs.
For piants with HPI shutoff head above the PORV set point, a feed and bleed system is acceptable. For plants with a lower HPI shutoff head, a bleed and feed system will be needed.
1 i
For either type of system sufficient sets of analyses need to be performed to:
(1) Detennine the number and capacity of the PORY valves needed for successful system operation, and (2) Develop detailed emergency procedures, including limitations on the earliest and latest times, after loss of all feedwater, that the bleed i
and feed (or feed and bleed) system can be initiated.
The analyses shall consider initiation of the system at various times after loss of all feedwater supply. Therefore, some sets of analyses will start 2/ Single failure requirement is applicable because of the safety injection function for core cooling, not because of the " feed" function for decay heat removal.
1 I
6
= =. -
,.~y_
~~, - -, ~..
..,,,_.....m.
...,,,-..r.,,..,,.,
,...,,,,, -,.,,,~.,,,,...,,
[
-. - ~ -
~ _.
ll 4
I 3
)
shortly af ter the feedwater supply is lost with tne water level in the pressur-izer at approximately its normal operating level. Other sets will start the i
analyses at a later time with the pressurizer full of water (due to formation l
of a steam bubble at the top of the reactor vessel).
Still other analyses will assume system operation is initiated after the water has drained from the f
pressurizer (in CE and Westinghouse-designed plants).
Where experimental flow rate data for the PORY for single phase and two phase I
flow is available, this may be used in the analysis.
If experimental data is 1
not available, then a suitable range of flow rates (to cover uncertainties in flow characteristics) must be included in each set of analyses.
i The effect of the piping between the pressurizer and the FORV(s) and of any a
piping downstream of the PORV's on the flow rate through the PORV system should t
l either be shown to be negligible or be considered in the analysis.
l The acceptance criteria to be used to judge the acceptability of the bleed I
and feed or feed and bleed system shall be that there is no uncovery of the i
core (.two-phase level) during the event, i
3.
Procedures J
l Emergency procedures for the use of the bleed and feed (or feed and bleed)
I system shall be developed. These procedures shall describe the use of the system and reflect any limitations on the use of the system determined by i
the analyses described in Item 2.
Specific diagnoses steps and equipment i
operability checks that are to be performed before initiating bleed and feed operation are to be included in the procedure.
In addition, plants having i
a.
I autocatic closure at low reactor coolant pressure of the block valve upstream of the PORY will have to specify procedures for manually bypassing this automatic function during bleed and feed operation.
4 Technical Scecifications Operability and surveillance requirements for the PORV(s) and block valve (s) l would have to be added to the technical specifications.
e S
w.
-e-
-w w
,,w-
..e, e-y-p,--.-
v-~-
y--
,y,,.
,rw,-
wr T
'mM-'r y i
-y-r-y-wyy9-p=+-
y w