ML20205J093
| ML20205J093 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/20/1986 |
| From: | Tierman J BALTIMORE GAS & ELECTRIC CO. |
| To: | Thadani A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20205J095 | List: |
| References | |
| NUDOCS 8601300097 | |
| Download: ML20205J093 (8) | |
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l B ALTIMORE GAS AND ELECTRIC CHARLES CENTER P. C. BOX 1475 BALTIMORE, MARYLAND 21203 JOSEPH A.TIERNAN Vict PntssotNT NUCLEAR ENtnar January 20, 1986 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 ATTENTION:
Mr. Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing-B
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Amendment
REFERENCES:
(a)
I&E Inspection Report 50-317/83-02; 50-318/83-02 (b) 1&E Inspection Report 50-317/84-16; 50-318/84-16 (c)
I&E Inspection Report 50-317/85-03 (d)
Letter from Mr. A. E. Lundvall, Jr., to Mr. R. W. Reid, dated April 10,1978 Gentlemen.
The Baltimore Gas and Electric Company hereby requests an Amendment to its Operating License Nos. DPR-53. and DPR-69 for Calvert Cliffs Unit Nos. l & 2, respectively, with the submittal of the proposed changes to the Technical Specifications.
CHANGE NO.1 (BG&E FCR 85-62)
Change pages 3/4 7-5b and B3/4 7-2 of the Unit No.1 Technical Specifications, and pages 3/4 7-5a and B3/4 7-2 of the Unit No. 2 Technical Specifications, as shown on the marked up copies attached to this submittal.
DISCUSSION This proposed change to the Technical Specifications would reduce the required value for delivered Auxiliary Feedwater (AFW) flow necessary to remove decay heat and lower the 0
Reactor Coolant System (RCS) temperature to less than 300 F (as stated in tle Technical Specification Bases). The narrative in the bases concerning the setting of the.
AFW flow control valves has been clarified, accordingly. Additionally, two related
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Surveillance Requirements have been changed. The first change corrects an obvious fjjl3pDOCK0500033, M,3
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J Mr. A. C. Thadani January 20,1986 Page 2 typographical error (Unit 2 only). The second incorporates the updated AFW flow rate value delineated in the bases.
These changes are necessary due to the modifications proposed under Facility Change Request (FCR) 34-1094. This FCR was written to allow the AFW automatic recirculation valve in the motor driven train to be overridden. In this condition the valve will be in permanent recirculation. The valve was originally installed to decrease the loading on the diesels by eliminating the continuous recirculation from the pump. As the AFW system was tested it became apparent that with automatic auxiliary feedwater actuation at low steam generator pressure, the recirculation valve and flow control valves were causing a flow instability problem.
Several NRC Inspection Reports address this References (a) and (b) provide a brief history of this phenomenon. Reference concern.
(c) provices a rnore current review, addressing potential corrective actions.
t Corrective actions for the AFW flow instability problem are relatively straightforward; however, their impact on emergency diesel generator loading has had to be closely studied. Placing the automatic recirculation valve in permanent recirculation without restricting delivered flow to the steam generators would cause the pump to require additional power.
i This increased load would exceed the maximum diesel generator load -
that has been dedicated to AFW, per our design.
Calculations were, therefore, performed to reanalyze the minimum required long term AFW flowrate. The result of these calculations, referenced in the safety analysis for FCR 84-1094, was a required flow of 300 gpm. This lower flowrate was attained by crediting initiation of charging flow (one pump) at 60 minutes, reanalyzing the RCS heat capacity, and applying ANS-3 1971 for decay heat loads (as permitted by 10 CFR 50, Appendix K). With this AFW delivered flow, all decay heat can be removed and the RCS cooled down to less than 300 F from normal operating conditions in the event of a total loss of offsite power.
This cooldown is accomplished within the six hour criteria of the Updated FSAR Section 1
10.3.2.
4 DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS i
This proposed change has been evaluated against the standards in 10 CFR $0.92 and has been determined to involve no significant hazards considerations, in that. operation of tne facility in accordance with the proposed amendment would not:
(i) involve any increase in the probability-or consequence of any accident previously evaluated in the Updated Final Safety Analysis Report.
With the automatic recirculation valve in permanent recirculation, adequate AFW flow is provided to meet the-Updated FSAR and Technical Specification heat removal bases.
Additionally, by imposing a maximum flow limit on the motor driven AFW pump, none of the equipment is loaded beyond specified design limits.
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Mr. A. C. Thadani -
January 20,1986 Page 3 (ii) create the possibility of a new or different type of accident from any accident previously evaluated.
No new accidents are created because equipment desiga limits luve not been exceeded. The possibility for secondary equipment failures leading to a loss of all AFW or station blackout has not been increased.
(iii) involve a significant reduction in the margin of safety.
Calculations performed indicate that 300 gpm supplied from the motor driven
/.FW pump is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the RCS temperature to less than 300 F from normal operating conditions in the event of a total loss of offsite power. The Updated FSAR references Loss of Feedwater (LOFW) and Feedline. Break (FLB) as undercooling scenarios. (FLB is not a design basis event for Calvert Cliffs, but it was analyzed for peak RCS pressure and 10 CFR 100 boundary dose calculations in association with the addition of automatic initiation of AFW). For the LOFW event, the present Updated FSAR assumes that the Operator initiates AFW flow at 600 seconds with no minimum flow specified.
The Operator, as stated in the bases, is assumed to be able to increase ~ or decrease AFW flow to that required by existing plant conditions. Feedline Break is analyzed for peak RCS pressure and credits no AFW flow until well after the peak pressure has occurred. Therefore, this event is not affected by the change in AFW flow. Incorporation of this Technical Specification change does not result in a significant reduction in the margin of safety for either of these events.
CHANGE NO. 2 (BG&E FCR 85-95)
Change pages 3/4 7-25,7-26 and B 3/4 7-5 of the Unit I and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
IMSCUSSION BG&E actively participates in the Snubber Utility Group Database. Discussions between utility members have proven helpful by providing constructive suggestions concerning the inspection and maintenance of snubbers. A discussion held during a recent snubber conference prompted a review of Calvert Cliffs' snubber Technical Specifications. This review revealed the need to clarify our Technical Specifications to properly delineate the importance of snubber " type" when cpplying inspection criteria.
This proposed change to the Technical Specifications provides clarification of the Surveillance Requirements for snubber visual inspections and functional tests. The Bases have also been clarified to explicitly refer to " type".of snubber as a guide in selecting-inspection sample populations.
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Mr. A. C. Thadani -
January 20,1986 Page 4 BG&E currently atilizes two types of hydraulic snubbers, both produc.ed by the same manufacturer (Grinnell). The two types operate similarly in principle, but have different design features (irrespective of capacity related dimension dif ferences). Small bore snubbers installed on various piping systems at Calvert Cliffs have bore sizes that range from 1 1/2" to 6". All small bores have the same design valve block; and, have only one valve block per snubber..The large bore snubbers, installed on the steam generators at Calvert Cliffs, all have 10" bores. The valve blocks on these snubbers are different in design than those on the small bore snubbers. Additionally, there are two valve blocks per snubber.
- The differences between these two types of snubbers is apparent upon a review of functional testing load and acceptance criteria. Small bore snubbers are designed for loads up to 72,000 pounds force.
The Locking Velocity (LV) and Bleed Rate (BR).
acceptance criteria are as follows:
Inches Per Minute (IPM)
LV: 1-40 (Adjusted for room temperature.)
BR:.25-25 The large bore snubbers are designed for loads up to 300,000 pounds force. Their acceptance criteria is much more restrictive:
Inches Per Minute (IPM)
LV: 1.25-1.75 BR: 0.0625-0.1875 Although all snubbers at Calvert Cliffs are currently produced by the same manufacturer, Surveillance Requirement 4.7.8.1 was expanded using standard industry
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phraseology to define " type" as "of the same design and manufacturer." inclusion of the l
word " manufacturer" precludes the necessity of a future Technical Specification change should Calvert Cliffs replace certain snubbers with a different make.
The words
" irrespective of capacity" were dropped from this standard definition so as not to confuse the issue of small bore capacity (same design) with large bore capacity (different design). This difference between small and large bore hydraulic snuSbers was explicitly.
referenced in the Technical Specification Bases to prevent any -future interpretive impasse.
The basing of snubber inspection populations on " type" of snubber design / manufacturer is not without precedent. The Combustion Engineering Standard Technical Specifications, along with the Technical Specifications of various operating utilities contacted, include guidance by " type" for the choosing of snubber inspection populations. BG&E consciously excluded " type" from its customized Technical Specifications since large bore steam generator snubbers wsre not required to be functionally tested in the past. The recent requirement to functionally test large bore snubbers has prompted this re-evaluation of snubber " type" in the Technical Specifications. BG&E feels snubber performance is best tracked by comparing operational data between like types.
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Mr. A. C. Thadani January 20,1986 Page 5 DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposed change has been evaluated against the standards in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or No modification to plant equipment is made by this proposed change. Each snubber " type" will continue to be inspected in accordance with standard acceptance criteria.
(ii) create the possibility of a new or different type of accident from any accident previously evaluated; or This proposed change does not involve a change to any component, system alignment, or operating procedure and would not create the possibility of a new or different kind of accident.
(iii) involve a significant reduction in a margin of safety.
This proposed change only clarifies the current Technical Specification wording by recognizing the two " types" of snubbers currently utilized at Calvert Cliffs.
This change allows for functional and visual snubber inspection criteria to be applied to each snubber " type" separately. This change does not involve a significant reduction in the margin of safety in that' the standard snubber acceptance criteria for each type are not being altered.
CHANGE NO. 3 (BG&E FCR 85-66)
Change pages 3/4 7-18,19, and 20 of the Unit Nos. I and 2 Technical Specifications as shown on the marked-up copies attached to this submittal.
DISCUSSION This proposal would make changes to the Control Room Ernergency Ventilation System Surveillance Requirements which perform the laboratory and in-place testing of the charcoal adsorbers and HEPA filters, and which verify control room isolation on a high radiation signal. These changes were discussed with the NRC previously as a result of a control room habitability re. view conducted in September 1985.
The proposed change to the charcoal adsorber laboratory analysis would change the required demonstrated efficiency from 90% to 97%; the temperature at which the analysis is done from 130 C to 30 C; and, the procedure which is used for the analysis from that in American National Standards Institute (ANSI) N510-1975 to that in American Society for Testing and Materials (ASTM) D3803. The required efficiency is
c.
Mr. A. C. Thadani January 20,1986 Page 6 proposed to be changed so that the laboratory analyses support the assumption of 9596 efficiency for removal of organic iodide by the adsorbers during the design basis radiological accident. This assumption was used in the Control Room dose calculation as stated in Reference (d).
Based on previous laboratory analyses, the demonstrated removal efficiency has not been less than 98.896 since system operation began. The
' temperature at which the laboratory analysis is conducted is proposed to be changed to 0
30 C so that the laboratory conditions more closely reflect the accident conditions to which the adsorber will be subjected. A 30 C laboratory analysis is more conservative than a 130 C analysis.
The procedure which is used for the laboratory analysis is proposed to be changed in response to NRC requests. The proposed laboratory procedure, ASTM D3803 is consistent with Regulatory Guide 1.52.
The proposed change to the in-place testing of the charcoal and HEPA filters would change the procedure used from -that in ANSI N510-197) to that in Regulatory Guide 1.52.
This change is admirustrative in nature and is proposed in response to NRC requests.
The proposed change to the surveillance which verifies control room isolation on a high radiation signal would clarify the wording to make the surveillance specifically address all appropriate isolation valves.
The proposal would not change the intent of the surveillance.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS This proposal has been reviewed against the standards set forth in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or The proposed changes to the Control Room Emergency Ventilation System
. Technical Specifications would make the surveillances more restrictive, or would make administrative and procedural changes that are consistent with NRC guidance (Regulatory Guide 1.52) and NRC verbal requests. Therefore, this proposal would not involve a significant increase in the probability or consequences of an accident previously evaluated.
(ii) create the possibility of a new or different kind of accident from any accident previously evaluated; or This proposal would not modify equipment design or operation and, therefore, would not create the possibility of a new or ditferent accident.
(iii) involve a significant reduction in a margin of safety.
Since the proposed changes are either more restrictive or in accordance with NRC guidance, no margin of safety will be significantly reduced.
t Mr. A. C. Thadani January 20,1986 Page 7 CHANGE NO. 4 (BG&E FCR 85-103)
Change page 3/4 9-15 of the Unit 1 and 2 Technical Specifications as shown on the marked-up copies attached to this transmittal.
DISCUSSION This proposal would change Technical Specification 4.9.12.d.2, which verifies that each exhaust fan in the Spent Fuel Pool Ventilation System maintains a negative pressure in the spent fuel storage pool area. Currently, Technical Specification 4.9.12.d.2 requires that a " negative pressuret 1/3 inches Water Gauge" be maintained by each exhaust fan.
This proposal would change the surveillance to require that a " measurable negative" pressure be maintained by each exhaust fan.
This change would provide greater flexibility in the performance of. this surveillance while still meeting its intent.
A mimimum a!!owable "measurabic negative" pressure will be estabilshed based on the sensitivity of the dif ferential pressure instrument used for the surveillance and will be controlled administrative!y. The differential pressure established will be verified by smoke test of area access doors and hatches to be negative enough to prevent any air leakage out of the spent fuel storage pool area.
DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS I
This proposal has been reviewed against the standards set forth in 10 CFR 50.92 and has been determined to involve no significant hazards considerations, in that operation of the facility in accordance with the proposed amendment would not:
(i) involve a significant increase in the probability or consequences of an accident previously evaluated; or The design basis for the Spent Fuel Pool Ventilation System is to ensure that l
all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere. The system is assumed to perform this function for a Fuel Handling Incident as mentioned in the Final Safety Analysis Report (FSAR),
j Section 14.18. A " measurable negative" pressure in the spent fuel storage pool area will continue to assure that air flow is into this area from outside.
Therefore, the amount of radioactive material released during a Fuel Handling Incident would not be significantly increased.
(ii) create the possibility of a new or different type of accident from any accident previously evaluated; or i
This proposal would not modify equipment or change system alignments. The procedural change involved would not create the possibility of a new or different accident.
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s Mr. A. C. Thadani January 20,1986 Page8 (iii) involve a significant reduction in a margin of safety.
This proposal ~ would allow fuel movement in the spent fuel storage pool area when pressure is less negative than 1/3 inch. Water Gauge. However, as discussed in (i) above, the amount of radioactive material released during potential incidents would not be significantly increased, so no safety margin would be significantly reduced.
SAFETY COMMITTEE REVIEW These proposed changes to the Technical Specifications and our determination of significant hazards have been reviewed by our Plant Operations and Off-Site Safety Review Committees, and they have concluded that implementation of these changes will not result in an undue risk to the health and safety of the public.
FEE DETERMINATION Pursuant to 10 CFR 170.21, we are including BG&E Check No. (1023626) in the amount of
$150.00 to the NRC to cover the application fee for this request.
Very truly yours, a
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- 3. A. Tiernan Vice President - Nuclear Energy STATE OF MARYLAND :
TO WIT:
CITY OF BALTIMORE :
Joseph A. Tiernan, being duly sworn states that he is Vice President of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belieft and that im was authorized to provide the response on behalf of said Corporation.
i WITNESS my Hand and Notarial Seal:
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Notary Public 1f/ff (
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My Commission Expires:
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