ML20245F728
| ML20245F728 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/18/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20245F720 | List: |
| References | |
| NUDOCS 8905020476 | |
| Download: ML20245F728 (10) | |
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UWITED STATES
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SAF,,g(EVALUATIONBYTHEOFFICEOFNUCLEARREACTORREGULATION d.
1 RELATED'T0_A3ENOMENJ NO.63T0 FACILITY OPERATING LICENSE NO. DPR-22
' NORTHERN STATES POWER COMPANY
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MONTICELLO NUCLEAR GENERATING PLANT DOCKET NO. 50-263
- 1. 0 INTRODUCTION 1
By letter dated May 5,1986, the Northern Stater Power Company (NSP or the licensee) prooosed changas to the Technical Specifications (TSs) appended to l
Facility Operating License No. DPR 22 for the Monticello Nucicar Generating Plant.
The proposed. amendment included changes.resulting from a detailed review of the TSs that occurred following the 1985. plant refueling end recircu-i lation piping. replacement outage.
Several of the proposed changes are adminis-trative in nature or are proposed to clarify the application of the existing TSs.
The specific changes and our evaluation of each change are presented below.
- 2. 0 DISCUSSION AND EVALUATION The T3 changes proposed by the licensee, including our assessment of the
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acceptability of each change, follows:
a.
Defirnition of Core Alteration
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Revise the definition of " Alteration of the Reacter Scre" in Section 1.0.A by 1
adding the words, "with the vessel head removed and fue) in the vessel," to the end of the first sentence.
4 According to the licensee, a literal interpretation of the existing definition would require movements to be corisidered core alterations even when fuel is not
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present within the vessel.
This is not necessary, and it creates conflicts with other Specifications.
For example, Specification 3.1>0.8, which uses this definition, would require SRM operability when no fuel is in the vessel.
We agree that the proposed change will clarify the definition of " Core Alteration."
b.
2.3 Bases Correction Delete the partial sentence in the first line of the first paragraph of the Section 2.3 Bases on page 17.
These words should have been delef A with a previous license amendment request, but were left through an oversight.
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Table 3.1.1, Startup Mode Operability Requi"enents 1
1 Move the reference to Note 4 from the " Refuel" collumn to the " Trip Function" column so the r.ote is applicable to all modes, and add the following new note to the table:
9.
High reactor pressure and main steam line high radiation are not required to be operable when the reactor vessel head is unbolted.
- dd a reference te Note 9 to the table entries for high reactor pressure and A.in steam line high radiation.
We agree with the licensee that the startup mode operability requirements listed for high drywell pressure, high reactor pressure, and main steam line high radiation are unnecessarily restrictive for activities such as low power physics tests and that it is desirable to eliminate such unnecessary requirements from the TSc.
Frr example, with the 9essel head upholt.ed, high reactor pressure and steam line radiation functions are not necessary.
High drywell pressure j
functions are not necessary when containment integrity is not required.
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Table 4.1.1_and Table 4.2,1, Variable Surveillance Frequencies, a,nd Associat & Bases Delete Note 1 of Table 4.1.1 and Note 1 of Table 4.2.1.
Delete Figure 4.1.1 i
.and correct the Lict of Figures to reflect deletion of this figure.
Delete all l
references to Note 1 on both tables and replace with a requirement for monthly
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surveillance.
Del e those portions of the 4.1 and 4.2 Bases which refer to 1
variable surveillant frequencies.
I hote 1 of these tables allows certain surveillance intervals to be lengthened up to a maximum of three months by application of Fiqure 4.1.1.
Lengthening of surveillance intervals is based on the number of unsafe failures that are experienced over a period of time.
Several years ago NRC asked the licensee not to use TS Figure 4.1.1 to lengthen surveillance intervals, and the licensee is reluctant to do so also since this would require periodic changes in test j
intervals and records.
As a result, monthly testing always has been conducted i
even though the T5s would permit less frequent testing.
Eecause this is a j
potentially confusing situation, we agree that the TSs should be revised as proposed to eliminate the option to extend surveillance intervals.
e.
SRM Not-Full-In Rod Block Interlock Conflicts i
Add a new Note 9 to Table 4.2.1 as follows:
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9.
Testing of the SRM Not-Full-In rod block is not required if the SRM detectors are secured in the full-in position.
Add a reference to Note 9 on Table 4.2.1 under item 8 of Rod Blocks. Change the item to read, "SRM Detector Not-Full-In Position" instead of, "... not in Start-Up Position." Change the sensor check requirement from " Note 2" to "None."
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l The existing testing requireinent for the SRM Not-Full-In rod block interlock conflicts with normal CRD maintenance work.
The specific appli1 cation of the j
test requirements could be inconsistent with the normal and prudent practices of rerouting the SRM cables to allow CR0 maintsaance and securing the SRM detectors in the full-in position.
Also, it is not ponible to perform the required sensar check of the interlock.
It is an on-off device, not an anahg signal subject to sensor checks, Because of the need to reroute the drive cables to allow normal CRD work and the lack of space under the vessel, it is preferable not to perform detector withdrawals for testing.
Such testing would damage the reconfigu ed cable.
Restoration of original cable configuration to allow testing would result in additional personnel exposure, additional wear, and risk of danage to cable and connector asseabties.
Verifying that the detectors are full-in and securing the detector drive powe; in "off" enforces the condition under which the
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interlock is satisfied.
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NRC inspectors have accepted this procedure in the past for the reasons stated, and as such, it is desirable to revise the TS wording to agree with this practice.
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Clarification of Containment _ Isolation Instrumentation Surveillance l
on Table 4.2.1
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l Expand the headings for main steam, HPCI, and RCIC isolation by adding a reference to the containent isolation group and add a new category for Group 2 and Group 3 containment isolation.
Delete Note 7 and all references tc Note 7 in the table.
Add a new Note 10 as follows:
10.
Uses contacts from scram system.
Tested and calibratea in accordance with Tables 4.1.1 and 4.1.2.
Add a reference to Note 10 for containment isolation Groups 2 and 3 reactor low water level and drywell high pressure surveillance.
We agree with the licensee that the existing TS surveillance requirements for containment isolation functions are misleading and imprecise.
The proposed changes would clarify and expand Table 4.2.1 to more clearly list the contain-ment isolation logic inputs and also note that some of the signals for Group 2 are derived from the scram logic.
Note 7 was added as part of an earlicr TS change accepted by the NPC.
The intent was to have Table 4.2.1 cover containment isolation surveillance.
The earlier change was not as precise as it should have been, and this further caange provides that precision.
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Bases for Specification 4.0 Revise the Bases section to exclain the surveillance testing requirements in Section 4.0 of the TSs and add information to assist in understanding and applying this section.
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~4-The proposed wording is derived from the NRC Standctd TSL Ad6 tional clarification of the surveillance interval tolerance is derived from clarifying information contained in NRC Inspection Report 50-263/85012(DRP) acted July 19, 1985.
J This addition to the Bases will help in understanding Section 4.0 of the TSs.
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This section was recently added to provide general requirements for the l
Surveillance Program. The proposed wording summarizes the application related to surveillance intervals and surveillance scheduling established with NRC inspectors over a period of many years.
h.
Table 3.2.5, Note 1, ATWS Instrumentation Requirements Revise Note 1 of Table 3.2.5 to read:
1.
When one of the two trip systems is made or found to be inoperable, restore the inoperable trip system to operable status within 14 days or place the plant in the specified required condition within the next eight hours.
When both trip systems are inoperable, place the plant in the specified required condition within eight hours unless at least one trip system is sooner made operable.
The existing Note 1 is inconsistent with the requirements for minimum number of operable or operating trip systems in laole 3.2.5.
A loss of one trip system would require a plant shutdown since it is rot possible to place a trip system in a tripped condition without actually causing actuation of the logic (this is 1 of 2 logic).
As long as the remaining tri; system is operable, this is an unnecessary requirement.
One operable trip system is sufficient to initiate the required protective action.
We believe this was an oversight on the part of the licensee when this TS change was first developed.
However, it should be noted that the staff is presently reviewing ATWS requirements to determine wheter, and to what extent, Technical Specifciations are appopriate.
The staff will provide guidance regarding the Technical Specification requirements for ATWS at a later date.
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Control Rod Accumulator Operability Clarification Delete the last paragraph of Specification 3.3.0 and redesignated items 1 and 2 under 3.3.0 as items 3.3.D.1(a) and (b).
Reword the opening paragraph as follows:
Contrcl rod accumulators shall be operable in the Startup, Run, or Refuel modes except as provided below.
Add Specification 3.3.0.2 as follows:
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In the Refuel Mode, a rod accumulator may be inoperable provided:
i (a) All fuel is removed from the cell containing the associated control rod, or
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(b) The one-rod-out refuel interleck for the associated rod drive is operable.
We agree with the licensee that a specific application of the last paragraph of this specification could preclude normal CR0 maintenance where the rod out refuel interlock is not bypassed and fucl is not removeG from the cell.
This could force cell unloading for all drive changeouts.
This was not the intent of the specification.
The specification was intended to apply only to the situation of multiple CR0 removal for extended core and control rod drive maintenance, controlled by Specification 3.10.E, where the rod out interlock is bypassed for withdrawn rods.
It is not impractical to unload fuel cells during actual refueling with the vessel head removed.
It is impractical, however, to require cell unloading for situations requiring drive maintenance after the core is fully reloaded and the head has been replaced.
Accordingly, the proposed change is acceptable.
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High Pressure Coolant Injection (HPCI), Automatic Pressure Relief System (APRS), and Reactor Core Isolation Cooling (RCIC) Operability Conditions Revise Specifications 3.5.D.1, 3.5.E.1, and 3.5.F.1 se that operability of these systems is not required above 350 psig during reactor coolant system leakage and hydrostatic tests by revising the Operability condition to read,
"..whenever the reactor pressure is greater than 150 psig and irradiated fuel is in the reactor vessel, except during reactor vessel hydrostatic or leakage tests." Also, reformat pages 109 and 110 to move the headingt for the APRS sections to the top of page 110.
We agree with the licensee that a specific application of the current require-ments for HPCI, APRS, and RCIC system operability conflicts with the requirement to perform reactor coolant system leakage tests following eac h refueling outage, and reactor coolant system hydrostatic tests at ten year intervals and following major system repairs or modifications.
The TSs currently recoire these systems l
to be operable when irradiated fuel is in the vessel and reactor pressure is greater than 150 psig.
Since these systems are designed to operate from a source of steam, they cannot be made operable 'during leakage or hydrostatic tests when the vessel is flooded and reactor coolant temperature is below saturation temperature.
i The NRC staff was contacted during the last refueling outage to obtain concur-rence with the logical understanding that these systems are in fact not required to be operable during leakage and hydrostatic tests.
The TSs for many Other beiling water reactor plants contain sfailar conflicts.
The proposed wording change climinates the need for this interpretation.
k.
Clarification of Primary Containment Requirements Reword Specification 3.7.A.1 as follows: "When irradiated fuel is in the reactor vessel and either the reactor coolant temperature is greater than 212 F or work is being done which has the poteritial to drain the vessel,, the follow-ing requirements shall be met except as permitted by Specification 3 5.G 4...."
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We agree with the licensee that a specific application of existing TS 3.7.A.1 would not allow draining of the suppression chamber when irradiated fuel was not in the reactor vessel and work which would, or had the pctential, to drain the reactor vessel was in progress.
This is due to the oinmission, in this case, of the standard wording, "when irradiated fuel is in the reactor vessel."
All other Monticello primary containment and ECCS TSs contain this provision.
1.
Rjactor Coolant System Venting and Requirements for Secondary Containment In Specification 3.7.C.2.b, delete the phrase, "...and the reactor coolant system is vented."
The requirement for the reactor to be vented at a condition for not requiring secondary containment conflicts with normal and reasonable activities during outages.
For example, reactor vents must be closed to perforsn vessel leakage and hydrostatic testing.
At other times, it is prudent to close reactor vents for radiological protection purposes.
There it, r.o basis for relating secondary < containment requirements to reactor venting.
Closing reactor vents during an outage when secondary containment was not established could be considered as a violation of the TSs even though the event would have no safety implications.
In addition, the standard TSs do not include this or similar limitations on reactor venting.
m.
Extended Core and CR0 Maintenance TS Conflicts Delete Specification 3.10.E.2 and redesignated Specifications 3.10.E.1 and 3.10.E.
Reword the first portion of the Specification to read, "More thar, one control rod may be withdrawn from the reactor core during outages provided that, except f or momentary switching to the Startup mode for interlock testing, the reactor inode switch is locked in the Refuel position _
The refueling interlock...."
i Change " withdrawn control rod" to " control rod" kn two locationt..
We agree with the licensee that the existing Specification 3.10.E.2 is totally redundant to Specification 3.10.B and therefore unnecessary and possibly confusing.
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i We also agree that there is a conflict between the existing Specification 3.10.E.1 and Specification 4.10. A and that a literal interpretation of the existing specification would prohibit rod withdrawal for normal operation and testing.
Specification 4.10. A requires weekly checks of the refueling interlocks until core alterations are completed and they are no longer required.
Core a3 tera-tions, as defined in Gection 1.0, occur throughout most of an outage.
Durirg thic time it is normal to have ccetrol rods removed from the core for mainten-ance in accordance with Specificatiori 3.10.L I.
The conflict is that Specifi-cation 3.10.E.1 requires the mode switch to be locked in " Refuel," but the weekly check of refueling interlocks requires switching momentarily to the "Startup" mode.
The iead-in statertent of tiie existing TS does not limit its applicability.
The current wording, taken literally,, would require that the mode switch be locked
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1 7-in refuel for any rod withdraw #1 from the core.
Tithough obviously not the intent, this would prohibiit more than one rod fron ever being withdrawn.
Accident Monitoring Instrurnentation Operabili"y Conditions n.
Revise Specification 3.1.4 to require operability of accident monitoring instrumentation, "...vhenever irradiated fuel is in the reactor vessel and reactor coolant water temperature is greater than 112 F...."
Revise the notes in Table 3.14.1 to require placing the plant in the c.old shutdown condition within 24 nours when required conditions of instrument openbility are not satisfied.
We agree with tne licensee that the existing vording for Specification 3.14 requires operability of this instrumentation in the startup and run modes.
During outages, the mode switch rnust often be placed in the startup position to perform tests or other normal operations.
However, acc:ident monitoring instru-mentation may be inoperable during an outage due to normal activities or conditions.
For example, SRV removal would render SRV position indicatior, inoperable and vessel draining would rander vessel level instrumentation
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There is no reason for accident instrumentation operability to be based on mode switch position.
A more desirable wording for the TS is to make it consistent with other accident mitigation system operability requirements (i.e., :bove 212 F) and the NRC Standard TSs.
Tables 3.14.1 and 4.14.1, Suppression Pool Temperature Monitoring o.
Iristrum?ntition Add the Suppression Pool Tengeratu.ce Monitoring instrumentation to Tables 3.14.1 and 4.14.1.
This pstrumentation was added as a result of the. Mark I containment Long-Term Improvement Program to accurately. monitor suppression peal average f.emperature.
We find that the <>perability and surveillance requirements proposed for this instrumentation are consistent with other TSs for accident rnonitoring i
ires trumentation.
-ihe SuppresC on Pool T4mperature Monitoring System (SPOTMDS) instal kd at Monticello was described in Volume 1, Section 5, of the Maticello iFt Unique Analysis Report submitted tb NRC on December 15, 1982.
This design m reviewed l
and approved oy the NRC, and a safety en.luation was issued on September 11, 1985.
In the ever.t of inoperability of SPOTMOS, alternate methcds Pf, monitoring suppression pool temperature are available until operability of SPOTM05 can tu MS'ared.
These methods fraclude use of an a? ternate multipoint recorder and tgerature sensors in the suppression poo't (the original temperature monitoring system).
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'CjarificationofTable4.14.1SensorChecks Provide additional notes in Table 4.14.1 to clarify sensor check requirements l.
for reactor water level, SRV valve position pressure switches, and SRV valve position thermocouple as follows:
(2) Once/ month sensor check will consist of verifying that the pressure switches are not tripped.
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(3) Once/ month sensor check will consist of verifying that the fuel zone level indicates off scale high.
(4) Following every Safety / Relief Valve actuation it will be verified that recorder. traces or computer logs indicate sensor responses.
Add a reference to Note 2 for SRV position pressure switches.
Add a reference to Note 3 for reactor vessel fuel zone water level, and add a reference to Note 4 for SRV position pressure switches and thermocouple.
We agree with the licensee that the specific application of the requirements for sensor checks of the SRV position pressure switches would require operation of the SRV's once per month.
It would also require establishing an abnormal reacto level to perform a sensor check of the fuel zone level instrument.
Table _14.1.1 notes have been revised to clarify the intent of these sensor checks, which is to require a verification of sensor operation following each SRV actuation, and once each month to verify that the fuel zone indicator is off scale high and the SRV pressure switches are not tripped.
These sensor checks do not comply literally with the definition of " Sensor Check" in Section 1.0 of the TSs, and it is important that notes to Table 14.1.1 cMarly specify the intended requirements.
To summarize, the proposed TS changes itemized above either improve clarity and logic, p.rovide some relief from some of the restrictions found to be either unnecessary or impracticable to perfmi, add a new requirement considered desirable by the NRC staff, and/or involve an administrative change, as follows:
Item a. - clarifies the definition of " Core Alteration."
Item b. - corrects a typographical error in the Section 2.3 Bases.
Item c. - corrects and clarifies the Startup Mode operability requirements for high drywell pressure, high reactor pressure, and main steam line high radiation.
Item d. - deletes the obsolete provision of the TSs which permits surveil-lance intervals to be extended.
Item e. - corrects conflicts with the SRM-Not-Full-In rod block interlock and CR0 maintenance.
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Item f. - corrects and clarifies the surveillance requirements for containment isolation instrumentation.
Item a. - adds a new section to the Bases explaining general surveillance requirements.
Item h. - corrects the action statements for ATWS instrumentation to correspond with the acceptable as-installed logic.
Item i.
c'.arifies CR0 accumulator operability requirements.
Item J. - corrects the HPCI, RCIC and APRS operability requirements to permit reactor coolant system leakage and hydrostatic testing.
Item k.
clarifies the requirements for containment integrity when no fuel is in the reactor.
Item 1. - corrects and clarifies the relationship between secondary containment requirements and reactor venting.
Item m. - clarifies the requirements for extended CRD maintenance.
Item n. - corrects and clarifies the operability conditions for facility monitoring instrumentation.
Item o. - adds TS LCOs and surveillance requirements for suppression pool temperature monitoring instrumentation.
Item p.
' clarifies the meaning of sensor checks for safety / relief valve position pressure switches and reactor fuel zone water level instruments-tion.
With the exception of Items b. and o., all of the changes have the intent of eliminating conflicts and problems in understanding the TSs.
These items were identified during a detailed review of the TSs by senior reactor operator licensed members of the Monticello technical staff.
This review was made to fulfill a commitment made to NRC management following the discovery during the 1985 refueling and maintenance outage of a number of conflicts in the TSs. W ile some relief frem impossible or unreasonable restrictions is granted ~1n several instances (e.g. HPCI will no longer be required operable during hydrostatic tests - but because the vessel is filled solid with subcooled water during these tests it is an impossible condition to impose), the requested changes will not, in any significant way, change the way the plant is operated or maintained.
Item o. adds new requirements for an instrumentation system instelled to meet the requirements of the NRC approved Mark I Containment Long-Term Program'and NRC Regulatory Guide 1.97, Revision 2.
This new and improved instrumentation system will enhance the information available to plant operators during normal and postulated accident conditions.
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3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves a change in the installation or use of a facifity component located within the restricted area as defined in 10 CFR Part 20 or changes an inspection or surveillance requirement.
We have determined that the amendment involves no significant 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 occupa-tional radiation exposure.
The Commission has previously published a proposed finding that this amendment involves no significant hazards consideration'and there has been no public comment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR j
51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The c+aff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
J. Stefano, NRR Dated:
April 18,1989
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