ML20086P602

Proposed Tech Spec 3/4.4.9 Re Pressure/Temp Limits to Modify Plant Heatup & Cooldown Curves & Max Allowable PORV Setpoint Curve for Cold Overpressurization Protection

ML20086P602
Person / Time
Site:	Callaway
Issue date:	12/18/1991
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20086P105	List: ML20086P103 ML20086P602
References
NUDOCS 9112270093
Download: ML20086P602 (19)

Text

._____

4 ULNRC-2537 Marked-ttp Technical Specification Paqog Page 3/4 4-29 Page 3/4 4-30 (Figure 3.4 -2)

Page 3/4 4-31 (Figuro 3.4-3)

Page 3/4 4-32 (Table 4.4-5)

Page 3/4 4-36 (Figure 3.4-4)

Page B 3/4 4-7 Page B 3/4 4-8 4

Page B 3/4 4-11 Page B 3/4 4-16

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3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldowr, celticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 100'F in any 1-hour period, b.

A maximum cooldown of 100*F in any 1-hour period, and c.

A maximum temperature change of less than or equal to 10*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and 'cooldown limit curves.

APPLICABILITY: At all times.

(

ACTION:

' With any of the abova' limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes: perform an enginsering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the React.or Coolant System remains acceptable for continued operation or be in at least h0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200*F and gyg 500 psig, respectively, w'ithin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, arid inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removeo and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H.;n ;;c;rdance with th; schedule in Tetle 4.4 L.

The results of these examinations shall be used to update l

Figures 3.4 ', 3.4-3, and 3.4-4.

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UPRATED TO 3565 MWt.

Figure 3.4-2 CALLAWAY - UNIT 1 3/4 4-30 Amendment No. 36

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Copper Content: 0.07 wt. %

1 Nickel Content: 0.59 wt. %

Nickel Content: 0.59 wt. %

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PR($$URE/ TEMPERATURE LIMITS (Continued) 2.

These limit lines shall be calculated periodically using methuca provided bhlow.

3.

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F.

4.

The pressurizer hestup and cooldown rates shall not exceed 100*F/h and 200*F/h. respectively.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583'F.

5.

Systen areservice hydrotests and in-service leak and hydrotests shall be perfer. 'd at pressures in accordance with the requirements of ASME Boiler and Pressu-1! Vessel Code.Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1972 Winter Addenda to Section l'1 of the ASME Boiler and Pressure Vessel Code.

I7 Heatup and cooldown limit curves are calculated using the most limitin value of the nil-ductility refarence temperature, RTHDT. at the ent' of :

fec-tive full power years (EFPY) of service life. The-9sEFPY tervice life period g

at ne 1/4T locadon in the con ngio is chosen such that the limiting RTN01 y

^

is greater than the RT of the limiting unirradiated material.

The selection HDT of such a limit.ing RT assures that all components in the Reactor Coolant NOT System will be operated conservatively in actordance with applicable Code requirements.

the reactor vessel materials have been tested to determine their ir.itial Reacter NOT; the results of these tests are shown in Table B 3/4.4-1.

RT operation and resultant fast neutron (E greater than 1 Mev) irradiation can Therefore, an adjusted reference temperature cause an increase in the RTNOT.

based v90n the fluence and copper content and phosphorus content of the material in question, can be predicted using Figure B 3/4.41 ed the largest value of aRI computed by either Regulatory Guice 1.99. Revision 2. " Effects of l

N07 Hesidual Elements on Predicted Radiation Darrage to Reactor Vessel.44terials."

or the Westinghouse Copper Trend Curves shorn in Figure B 3/4.5-2.

The heatup and cooldown limit curves of Figures 3.4-2 und 3.4-3 include predicted adjust-f-jFPYaswellasadjustmentsfor l

ments for this shift in RT at the end HDT

/7 oossible errors in the pressure and temperature sensing instruments.

Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185 are

.iva l l ab l e.

Capsules will be removed in accordance with tne requirements of ASTM E185-73 and 10 CFR Part 50. Appendix H.

%: = ;;i!4 ace :peci e*

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CAltAWAY - UNIT 1 B 3/4 4-7 Amendment flo. 36

l REVIslo'N J

REACTOR COOLANT SYSTEM a

-,g

, BASES i

FRESSURE/ TEMPERATURE LIMITS (Continued)

h

u

b ;h: a '- M k e."-0.

The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation darrage to the reactor vessel material by using the lead f actor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ARTNDT NDT for the equivalent capsule radiation exposure.

Allowable pressure temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section !!! of the ASME Boiler and Pressure Vessel Code as required by Appendix G to.10 CFR Part 50, and these methods are discussed in detail in WCAP-7924 A.

The general rnethod for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic f racture mechanics (LEFM) technology, in the calculation procedures a semi-elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section 111 as the reference flaw, amply exceed the current

)

capabilities of inservice inspection techniques. Therefore, the reactor

'j operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for. protection against nonductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature RTHDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approech for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, k, for the combined thermal and pressure stresses at any time during bestup g

or cooldown cannot be greater than the reference stress intensity factor, K;g, f or the r tal temperature at that time.

K is obtained from t.ie reference IR fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation:

IR = 26.78 + 1.223 exp [0.0145(T-RTHDT + 160))

Q)

K Where: K is the reference stress ' intensity factor as a function of the 'netal jp

Thus, temperature i and the metal nil-ductility reference temperature RTNDT.

CALLAWAY - UNIT 1 B 3/4 4-8

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T REAC10R (,00 TANT SYSTEM BASES COLD OVERPREs5U1E (Continued)

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possible by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; 3) instrument uncertainties; and 4) single f ailure. To ensure mass and heat input transients more severe than those assumed cannot occur, technical specifications require lockout of both safety iniaction pumps and all but one centrifugal chrging pump while in HODES 4, 5 and is with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50'F above primary temperature.

Exceptions to these mode requirements are acceptable as described below.

Operation above 350'F but less than 375'F with only one centifugal charging pump OPERABLE and no safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

As shown by analysis LOCA's occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump 6nd a single RHR pump with no credit for accumulator injection. Given the short time duration that the condition of having only one centrifugal charging pump OPERABLE is allow'ed and the probability of a LOCA occurring during this

, time, the failure of the single centrifugal charging pump is not assumeL Operation below 350*F but greater than 325'F with all centrifugal charging and safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During low prest,ure, low temperature operation all automatic safety injection actuation signals except Containment Pressure - High are blocked.

In norm 61 conditionF a

i. ingle failure of the ESF actuation circuitry will result in the starting of at

.)

most one train of safety injection (one centrifugal charging pump, and one safety injection pump).. For temoeratures above 325'F, an overpressure event occurring as a result of starting two pumps can be successfully mitigated by operation of both.

PORV's without exceeding Appendix G limit. Given the short time duration that this condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORV is not assumed.

Initiation of both trains of safety injection during this 4-hour time frame due to uperator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents.

Although COMS is required to be OPERABLE when RCS temperature is less than 3f.# F, operation with all centrifugal charging pumps and both safety injection pumps OPERABLE is acceptable when RCS temperature is greater than 35C'F. Should an inadvertent safety injection occur above 350*F, a single PORV has sufficient capacity to relieve the combined flow rate of all pumps. Above 350'F, two RCP and all pressuriter r,afety valves are required to be OPERABLE. Operation of an RCP climinates the possibility of a 50'F dif ference existing between indicated dhd dCtual RCS temperature as a result of heat transport ef fects. Consirkring instrument uncertainties only, an indicated RCS temperature of 350*F is 'suffi-ciently high to allow full RCS pressurization in accordance with Appendix G limitations.

Should an overpressure event occur in these conditions, the pres-Surizer safety valves provide acceptable and redundant overpressure protection.

The Maximum Allowed PORV setpoint for the Cold Overpressure Hitigation System will be updated based on the results of examinations of reactor vessel.

material irradiation surveillance specimens performed as required by 10 CFR N3 Part 50, Appendix H,rd '; ::: rd:n:: with the-echeduh '- Tek *

"5,

.j CALLAWAY - UNIT 1 B 3/4 4-16 e

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C

ULNRC 2537 SAEHILIWA1AMTlON This Technical Specification amendment incorporates information gained from sLcveillance capsule Y which was removed during Callaway Refuel 4.

Capsule Y is the second capsule to be removed f rom the reactor vessel in the continuing surveillance program which monitors the effects of neutron irradiation on the Callaway Plant reactor vessel materiale under actual operating conditions.

The modification of curves found in Figures 3.4-2, 3.4-3, and 3.4-4 is to incorporate the RT as determined from capsule Y.

Table 4.4-5 is being removed fhbb technical specifications in accordance with Generic Letter 91-01.

The Bases are revised to 1

reflect the change in the service life, the removal of Table 4.4-5, and to correct a typographical error in Table B 3/4.4-1.

The fluence calculation for this amendment is based on the Westinghouse version of the discrete ordinate code DOT-3.5 which uses the SAILOR cross section data set.

An R-theta calculation was performed with 18 groups above 1 Mev and S order of angular n

quadrature.

Scattering is treated with a P approximation.

The 3

fission energy spectrum accounts for anticipated changes due to burnup including increases in plutonium.

See WCAP-12946 (submitted to NRC by ULNRC-2461, 8/14/91) for a more complete di; cur:sion of the radiation analysis and neutron dosimetry methodologier.

Tha heatup and cooldown curves (Figures 3.4-2 and 3.4-3) were calculated in accordance with 20CFR50, Appendix G and ASME Code, Secticn III, Appendix G requirements using the most limiting value of RT for the reactor vessel.

The curves are applicable for the firgpT17 EFPY of operation and include margins of 10 degrees F and 60 psig to allow for possible instrumentation errors.

The changes to these curves are consistent with the desigt. basis as described in the Bases section of the Technical Specifications These curves are shifting up and to the left.

This is occurring because a less stringent methodology (Reg.

Guide 1.99, Rev. 2 versus Proposed Rev. 2) and actual fluence levels have been used in the calculations.

The PORV Setpoint Curve (Figure 3.4-4) for Cold Overpressure Mitigation System (COMS) is based on the heatup and cooldown curve s.

Therefore, anytime the heatup/cooldown curves are revised, the PORV Setpoint Curve and the COMS setpoints must be evaluated and revised if necessary.

For the purpose of this submittal, the heatup/cooldown curves were revised such that a revision to the PORV Setpoint Curve was required.

This curve shifts upward due to the changes as discussed above.

Page 1 of 3

t UhNRC.2537 The selection of the pressure setpoints for the PORV's is based on the use of nominal-upper and lower pressure limits specified by Appendix G to 10CFR50 and the reactor coolant pump shaft

- seals, respectively.

From the standpoint of determining the

- maximum Petpoint and proximity to Appendix G, the mass input mechaninma considered in the analysis involved the operation of a l

single charging pump at maximum flow with inadvertent isolation of letdown flow and RHR suction relief valves.

Inadvertent actuation of a safety injection pump was not explicitly analy cd j

since its operation is prevented by the Technical Specifications.

The-heat input mechanism considered for analysis involved a RCS

-coolant pump startup in one loop with a temperature asymmetry in the RCS, whereby the steam generator is at a temperature 50 de?rees F higher than the rest of the RCS.

The magnitude of the tu..perature difference would normally depend on the previous operation of the plant which allowed the asymmetry to develop.

However, 50 degrees F is considered a realistic maximum i

temperature difference and is also controlled by_ Technical Specifications.

1 Correcting the typographical error in Table B 3/4.4-1 is an enhancement to the current technical specifications which could serve to avoid confusion in the future.

- The proposed changes to Technical Specification 3/4.4-9 do not involve an unreviewed safety question because operation of Callaway Plant with this change would not:

1.

Increase the probability of occurrence or the consequences t

of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.

This change is based upon methods of specimen analysis which have been utilized for Callaway and other plants and which do not change the probabilities of accidents or the malfunction of any equipmen 2.

Create a possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report.

There is no new type of accident or malfunction being created and the method and manner of plant operation remains unchanged.

This change is based upon NRC and Westinghouse methodologies which do not impact the-current accident analysis as discussed in the FSAR.

3.

Reduce the margin of safety as defined in the basis for any technical specification. - This is based on the fact that the current-practices and procedures-for plant operation will not change.

Page 2 of 3 w

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ULNT.C 2537 The specimen withdrawal schedule (Table-4.4-5) is removed from Technical Specifications in accordance with the recommendations of.NkC Generic Letter 91-01.

This table will be naintained in the Callaway FSAR and will be included in the next revision of this document.

The removal from the technical specifications of the schedule for the withdrawal of reactor vessel material surveillance specimens will not result in any loss of regulatory j

control becaue changes to this echedule are controlled by the requiremente (

10CFR50, Appendix H.

i The proposed removal of Table 4.4-5 from Technical Specifications and the modification to the Bases to reflect this removal does not involve an unreviewed safety question becauco operation of Callaway Plant with this change would not 1)

Increase the probability of occurrence or the consequences l

of an accident or malfunction of equipment important to safety previously evaluated ir, the safety analysis report.

This change is based upon NRC Generic Letter 91-01.

Removal of this table from technical specifications is purely r

administrative and does not impact any accident or equipment.

2)

-Create a possibility for an accident or malfunction of a different type than any previously evaluated in the safety analysis report.

There is no new type of accident or malfunction being created and the method and manner of plant operation remains unchanged.

This change is administrative in nature only.

3)

Reduce the margin of safety as oefined in the basis for any technical specification.

This is based on the fact that the current. practices and procedures for plant operation will not change due to the administrative nature of this change.

Given the above discussions as well as those presented in the Significant Hazards Evaluation, the proposed changes do not i

adversely affect or endanger the health or safety of the general public or involve a significant safety hazard.

5 L

Page 3 of 3

ULNRC-2537 SIGNIFICANT 1{AZARD EVALUATION i

This amendment application is requested to incorporate information gained from surveillance capsule Y which was removed during the Callaway fourth refueling.

Capsule Y is the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Callaway reactor pressure vessel materials under actual operating conditions.

The modifications of the plant heatup and cooldown limitation curves and PORV Setpoint Curve (Figures 3.4 2, 3.4-3, and 3.4-4) are to incorporate the revised RT The revision to Bases Table B 3/4.4-1 is editorial in nkkdr.e only.

1.

This change does not involve a significant increase in the probability or consequences of an accident previously evaluated This amendment merely ensures that the acceptable range of operation is clearly defined using conservative and validated data obtained from the analysis of surveillance capsule Y and previously utilized methods of specimen analysis.

2.

The change does not create the possibility of a new or different kind of accident from any previously evaluated.

The;e is no new type of accident or malfunction being created and the method and manner of plant operation remains the same.

This change is based upon NRC and Westinghouse methodologies which do not impact the accident analysis.

3.-

This change does not involve a significant reduction in a margin of safety.

This is based on the fact that the change only revises the heatup and cooldown curves to reflect operational parameters based on surveillance capsule data.

The recalculated limit curves have the ma~a otgree of i

conservatism as the original curves, since they are based on the most limiting values of the nil-ductility reference temperature which includes tha radiation induced shift (change in RTNDT) as determined by the surveillance capsule analysis The specimen withdrawal schedule is being removed from the Technical Specifications.in accordance with the recommendations of NRC Generic Letter 91-91.

This table will be maintained in the Callaway FSAR.

The removal of this table trom the technical specifications will not result in any loss of regulatory control because changes to this schedule are contro]1.rd bi the requirements of 10CFR50, Appendix H.

Page 1 of 2

~.-

. ~.

l e

i ULNRC-2537 1.

Thie change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

This change is administrative in nature only.

j 2.

The change does not create the possibility of a new or different kind of accident from any previously evaluated.

There is no new type of accident or malfunction being created and the method and manner of plant operation remains the same.

3.

This change does not involve a significant reduction in a margin of safety.

This is based on the fact that the specimen withdrawal schedule is controlled by the requirements-of 10CFR50, Appendix H.

The removal of this table from technical specifications is purely administrative and does not-impact any margin of safety.

Based on the above discussions, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated or create the possibility of a new or different kind of accident-from any previously evaluated.

These changes do not result in a significant reduction in a margin of safety.

Therefore, it has been determined that the proposed changes.do not involve a significant hazards consideration.

s Page 2 of 2

?

s ULNRC-2 5 3 7 Environmental _ Consideration This amendment application is requested to incorporate information gained from surveillance capsule Y which was removed during the Callaway fourth refueling.

Capsule Y is the second capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the Callaway reactor pressure vessel materials under actual operating conditions.

The modifications of the plant heatup and cooldown limitation curves and PORY Sotpoint Curve (Figures 3.4-2, 3.4-3, and 3.4-4) are to incorporate the revised RT The revision to Bases Table B 3/4.4-1 is editorial in nNEdr.e.

The proposed amendment involves changes with respect to the use of facility components located within the restricted area as defined in 10 CFR Part 20, and changes a surveillance requirement.

Union Electric has determined that the proposed amendment involves no significnnt increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criteria for categorica.1 exclusion set forth in 10 CFR 51.22 (c) (9).

Pursuant to 10 CFR 51.22 (b) no environmental impact statement or environmental assessment need be prepared in connectios with the issuance of this amendment.

.