ML19289G309
| ML19289G309 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/02/1975 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19289G307 | List: |
| References | |
| NUDOCS 7908220440 | |
| Download: ML19289G309 (4) | |
Text
.
e t)V f
b*
j ZZACTOR COOLANT'SYSTDi OVERPRESSURE PROTECTION DURING STAEIUP CR ShdDTD I.
BACKCROUND The general design criteria (Appendix A to 10 CFR 50) specify in General Design Criterion (GDC) 15 regarding the reactor coolant system design that:
"The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operction, including anticipated operational occurrences."
Appendix A of 10 C7R 50 contains the following definition.
" Anticipated operational occurrences mean those conditions of nor=al operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pu=ps, tripping of the, turbine generator set, isolation of the main coidenser, and loss of all offsite power."
This me=orandum is intended to provide guidance to the RSB reviewers by clarifying the intent of the above criterion and definition as,
they relate to the potential overpressurization of the reactor coolant system during startup or shutdoun.
Inadvertent overpressuri:stion when the reactor coolant system is water-solid (startup or shutdown conditiens) could result from a ralfunctica or operator error.
The cencern is that the reactor coolant system (KCS) pressure =ight exceed the pressure-te=perature limitations on the reactor ves'sel for protection against brittle fracture.
The pressuru-temperature limits imposed on the reactor coolant pressure boundary (PI?B) daring operation and tests are provided to assure adequate safety margins against nonductile behavice or rapidly propagating failure of ferritic co ponents the RCFB, as required by GDC 31.
Appendices C and H cf 10 CFR 50 describe the conditions that require pressure-temperature limits and provide the general bases for these 14n<ce.
These appendices specifienlly require that pressure-te=perature limits
- cust provide safety =argins at 1nant as. great as those recot= ended in the ASME Boiler and Pressura Vessel Code,Section III, Appendix G.
s$
J
" Protection Against Nca-ductile Tailure/' during heatup, cooldown, and test conditions. Appendix G of the** f.:de provides a basis for dstermining allovabic pressure-tc=perature relationships for normal and upset conditions.
Section III - Division 1, Subsection NB, subparagraph NB-3113.2 of the Code defines upset conditions (incidents of moderate frequency) as follows:
"Any. deviations from Nor=al Conditions anticipated to occur often enough that design should include a capability to withstand the conditions without operational i= pair =ent.
The Upset Conditions include those transients which result from any single operator error or control malfunction, transients caused by a fault in a system co=ponent requiring its isolation from the system, and transients due to loss of load or power."
From this definition of Upset Conditions, the pressure transient while water-solid is clearly an upset condition.
Since 1970 there have been at least seven incidents in uhich the reactor coolant syste: pressure exceeded the pressure-teeperature limitations on the reactor vessel for protection against brittic fracture.
Table 1 attached su==arizes the cause, the exte'nt of the transient, the precourc li=it for the existing vessel temperature as specified in the plant technical specifications, and the length of ti=c the trcusient proceeded until it was ter=inated.
Standard Revicu Plan 5.3.2 which deals with the review of the pressure-temperature li=its, states that the pressure-temperature limits for normal operations (including startup and shutdown conditions) as given in the plant technical specifications are applicable to upset conditions.
Normal operating procedures must per=it variations frem intended cperation, including all upset conditions, without exceeding the limit curves.
II.
Review Objectives The objective of the RSD review of the potential overpressuri:stion of the RCS during startup or shutdown is to assure that the pressure-ce=perature limit curves for upset con'ditions are not execcded.
To satisfy the objective, the 353 review should include but not be limited to the following:
1.
An analvsis of the RCS pressure as a function of time for at least 600 ceconds while the RCS is vater solid. Tac analysis should assu=e the =ost li=iting conditions (e.g., one RIIR train operating or available for letdown, other components in nor=si operation then the cystem is water solid such as,pressuri:cr henters and charginc pucps, and-one or more reactor coolant pumps in operation) with the worst single failure or operator error.
Justification should be a
2043 040
\\
.r w 15 h(
provided~ for the choice of li=1 ting conditions and worst single failure or operster error used in the analysis.
2.
The capacity of any relief valves attached to the RCS a-.
a function of pressure should be provided as a part of the
, analysis. The analysis should show the RHR relief valve flow as a function of tice until the isolation valve (s) autom.atically close due to the high RCS pressure.
3.
If the review and analysis show that the RCS execeds its pressure-tenperature li=itations for nor=al and upset conditions, design changes should be required in order to mitigate the consequences.
Consideration might be given
~
to (1) increasing the capacity of RHR suction relief valves, (2) providing dual range relief / safety valves on the RCS pressurizer, and (3) installation of relief valve (s) on the discharge of the high pressure charging pu=ps that are valved into service du ing startup or shutdot.n.
0
.0 20.43 041 a
e 4
W 9
e
TA!!LE 1 INCIDENTS OF PRESSURE s
TRANSIENTS 11EYOND TECll. SPEC. LIMITS
\\
PRESSURE TRANSIENT TECH. SPEC.
TIME W INCIDENT CAUSE FROM,_(PSIC) TO LIMIT (PSIC)
TERMINATION (HIN,)
1.
Testing of RHR isolation valve automatic closure interlock 95 1300 450 18 fpolated letdown 4
2 Operator closed RilR isolation
/
values to prepare for maintenan'ce 100 1100 a 550 10 on RilR relief valvas 3
Operator error in failing to terminate charging pump operation 400 1290 530 Note 1 4
Oporator inadvertently isolated letdown 420 680 500 Note 1 5
_ Upk_ noun 420 670 500 3_
6 Oporator inadvertently isolated letdown 420 650 500 2_
7 Reactor coolant pump starting sucpt cold water thru hot sten a generator - pressure 400 640 500 1
increase due to thermal
_
- expansion
_a N
Cliote 1 - The available abnormal occurrence report does not provide this information.
DPET C
~
N
.