ML20138R239

Forwards Response to Five Open Items Identified in Technical Evaluation Rept on Util Response to Generic Ltr 82-33 Re Implementation of Rev 2 to Reg Guide 1.97 Concerning Emergency Response Capability

ML20138R239
Person / Time
Site:	FitzPatrick
Issue date:	12/24/1985
From:	Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Muller D Office of Nuclear Reactor Regulation
References
RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, JPN-85-91, NUDOCS 8512310089
Download: ML20138R239 (12)

Text

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,. 123 Main Street Wivte Rains, New York 10601 914 681.6240 N E Seni r Vice esident '

1# Authority "- ' o-"

December 24, 1985 JPN-85-91 l

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Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

) Attention: Mr. Daniel R. Muller, Director j BWR Project Directorate No. 2 Division of BWR Licensing

Subject:

James A. FitzPatrick Nuclear Power Plant i Docket No. 50-333 Emergency Response Capability - Conformance

, to Regulatory Guide 1.97 Revision 2

References:

1. NYPA letter, J. C. Brons to D. B. Vassallo, dated November 30, 1984 (JPN-84-77 ) regarding response to Generic Letter 82-33.

2. NYPA letter, J. C. Brons to D. B. Vassallo, dated June 28, 1985 (JPN-85-53) regarding revised schedule for implementation of Regulatory Guide 1.97.

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3. NRC letter, D. B. Vassallo to J. C. Brons, dated November 5, 1985 transmitted EG&G preliminary Technical Evaluation Report regarding the same subject.

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Dear Sir:

In Reference 1, the Authority responded to Section 6 of Generic Letter 82-33 and described plans and schedules for implementing Regulatory Guide 1.97 Revision 2 at the l-FitzPatrick Plant. Exceptions to certain of the guide's regulatory positions were justified in that letter. The implementation schedule was revised and updated in Reference 2.

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Reference 3 transmitted a preliminary Technical Evaluation Report (TER) documenting the NRC's review of the Authority's proposed plans. The TER identified five open items, requested a response and solicited the Authority's comments.

Attachment 1 replies to the five open items identified in the TER. Plans and schedules for two Regulatory Guide 1.97 variables, deferred until after the Authority had responded on the new rule on Anticipated Transients Without Scram (ATWS), are included as Attachment 2. Authority comments on the TER are included in Attachment 3. Attachment 3 also corrects the implementation schedule included with Reference 2.

If you have any questions regarding this information, please contact Mr. J. A.-Gray, Jr. of my staff.

Very truly yours,

-W u John . Brons Senior Vice President Nuclear Generation cc: Office of the Resident Inspector U. S. Nuclear Regulatory Commission P. O. Box 136 Lycoming, New York 13093

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l I-ATTACHMENT 1 TO JPN-85-91

New York Power Authority ,

James A. FitzPatrick Nuclear Power Plant l Docket No. 50-333 1

! DPR-59 l

> i NYPA Response to NRC letter of November 5, 1985

and the Technical Evalution Report regarding conformance to Regulatory Guide 1.97 Revision 2 4

1 NRC Comment 1: Neutron flux -- the licensee's present I instrumentation is acceptable on an interim basis i until Category I instrumentation is developed and l installed (Section 3.3.1).

l NYPA Response: The Authority will continue to use the existing neutron flux instrumentation.

. The Authority will follow the industry development j of new neutron flux monitoring equipment.

Replacement equipment will be evaluated when it

', becomes avaliable. The Authority will consider installing new' neutron flux monitoring j instrumentation at that time.

! NRC Comment 2: Radiation exposure rate -- the licensee should show 4

that the ranges supplied for this variable encompass the radiation level at the instrument location (Section 3.3.4).

! NYPA Response: (See response to NRC Comment 5 below.)

l NRC Comment 3: Standby liquid control system storage tank level --

i environmental qualification should be addressed in l accordance with 10 CFR 50.49 (Section 3.3.7).

1 NYPA Response: As the Authority stated in Reference 14, there are no components in the Standby Liquid Control (SLC) system which require qualification under the i provisions of 10 CFR 50.49 (Environmental Qualification of Electrical Equipment Important to j Safety for Nuclear Power Plants).

The SLC system design basis assumes the need for an j alternate method of reactivity control without a

! concurrent LOCA or high-energy line break. The environment in which the SLC tank ~ level l instrumentation must work is therefore a mild Page 1 l

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I e environment for equipment qualification purposes.

Therefore, no changes to the existing SLC tank level i instrumentation are necessary to comply with the l requirements of 10 CFR 50.49.

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. NRC Comment 4: RHR service water flow -- the licensee should either

! show that the range is adequate or provide the recommended range (Section 3.3.8).

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NYPA Response: An error was made in the Position Summary Table

- included with Reference 1. The table entry associated with RHR Service Water System Flow (Items A9 and D22-B) should have listed an installed instrument range of zero to one hundred and fifty c

percent (0% - 150%) of design flow. The installed

instrument range (0% - 150%) exceeds the range required by the regulatory guide (0% - 110%).

i Therefore, no plant modification or futher justification is necessary.

NRC Comment 5: Reactor building or secondary containment area radiation -- the licensee should show that the ranges supplied for this variable encompass the 4

radiation level at the instrument location (Section l 3.3.10).

l 1 NYPA Response: The existing Area Radiation Monitoring System (ARMS) j

consists of thirty monitors located throughout the '

l plant. Most instruments have a maximum indication of j 1 R/hr. ARMS for the refueling floor and chemistry

! laboratory have ranges of 1000 R/hr and 0.1 R/hr, respectively. Alarm points range from 1 to 200

mR/hr, depending on detector location.

I i The Authority considers the existing Area Radiation

Monitoring System to be adequate for the following l reasons

1. The Authority has previously stated that it is not necessary to enter the FitzPatrick Reactor Building following;an accident (Reference 7).

As a part of the our efforts to respond to NUREG-0578, Item II.1.6.B, and NUREG-0737 Item I II.B.2, the Authority performed a study to predict radiation levels in the plant in the vicinity of systems that may, as the result of an accident, contain highly radioactive material. This report identified vital areas and equipment in which personnel occupancy may be unduly limited or safety-related equipment Page 2

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1 j unduly degraded during post-accident operation.

> This study used the conservative assumptions of

, a post-accident release of radioactivity

equivalent to that described in Regulatory l Guide 1.3 (i.e., the equivalent of fifty

{: percent of the core halogens, one hundred i' percent of the core noble gases, and one i percent of the core solid fission products are contained in the primary coolant). The study considered only systems outside primary containment which may contain primary coolant or gases: HPCI, LPCI, RHR, Core Spray, l

Containment Air Dilution (including torus /drywell purge and vent system), Standby Gas Treatment and reactor coolant sampling lines. The following areas were considered in

this analysis

Control Room, Technical Support

!- Center, Primary Coolant System Sampling i Stations, motor control center locations, 2 electrical control panel locations, manual 4 isolation and control valve locations for vital '

equipment, Chemistry Laboratory and Counting l Room, areas where emergency maintenance may be

! desirable, and other general access areas.

j Twenty detector points were modeled in the 4 i

analysis.

4 The results of this study show that twenty-four

! hours after the accident, more than one-half of l the detector points modeled would be exposed to dose rates less than 1 R/hr (i.e. on-scale for

! most of the installed area radiation monitors).

l' After four days, Reactor Building doses rates were predicted to range from a maximum of 38 l R/hr to zero R/hr (3.7 R/hr average). After I fourteen days, the maximum dose rate decayed to

! less than 20 R/hr with an average of 1.7 R/hr.

The NRC's Office of Inspection and Enforcement conducted an inspection (Reference 9) in March of 1983 to review our post-accident shielding design review. As part of that inspection (Section 2e, Vital Area Accessibility Procedure Review), the inspectors reviewed three procedures that would be used post-LOCA. The inspectors determined that these three procedures (1) could be completed from the control Room, (2) contain appropriate provisions to assure controlled access to vital areas for post-accident operations, and (3) post-accident doses to plant personnel would be within the guidelines of NUREG-0737. Overall, i

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3 i our shielding design review was found to be consistent with the guidelines of NUREG-0737.

j The use of the existing area radiation monitors j in classifying emergencies is described in NYPA procedure IAP-2, " Classification of Emergency i Conditions."

l The source terms used in the post-accident shielding analysis are very conservative.

i Studies of the accident at TMI-2 confirm this (Reference 16).

t While some of the existing area radiation monitors will read off-scale during the early phases of the accident and recovery, this is acceptable since entry.into the Reactor

. Building is not required in the short or medium j term.

l 2. An extended-range area monitoring system would i not, by itself, provide adequate information

, during an emergency situation. Local

radiological surveys and airborne contamination

samples would be required prior to any entry I into the Reactor Building regardless-of whether

or not extended-range area radiation monitors were installed.

3. Portable radiation survey instruments are ,

adequate as a primary source of information considering that Reactor Building entry is not necessary in the short or medium term. These j instruments are' maintained and used as part of j normal plant operations. Plant personnel are i trained to use this equipment and are familiar

! with it.

4. In response to an NRC Emergency Preparedness l Appraisal (Reference 10, Appendir B, Item 11),

the Authority performed an engineering study on upgrading the existing Area Radiation Monitoring System. This study found an upgrade not to be cost' effective.

i During a subsequent IEE inspection (Reference 11), Han NRC inspector reviewed the study and it's conclusion. As a result, the inspector closed the open item.. i

5. The Authority does not consider area radiation monitors a feasible or desirable way of Page 4

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A detecting a breach of primary containment.

Primary containment breaches are better detected by plant noble gas effluent monitors.

i The BWROG has adopted this same position q (Reference 12).

In general, Reactor Building radiation exposure

rate will reflect the levels in the drywell I atmosphere and ECCS piping (reactor coolant).

Area radiation readings could be distorted as a result of direct shine from piping and i equipment carrying drywell atmosphere or i- reactor coolant. Local radiation exposure rate i

monitors could indicate ambiguous exposures due

, to the amount and arrangement of piping, the

! quantity of electrical penetrations, hatches and their scattered arrangement.

! 6. High-range radiation monitors have been j installed in other parts of the plant in

response to other NUREG-0737 items.

l The existing High-Range Effluent Monitori. ,

i System (HREMS) consists of three noble gas

monitoring units. HREMS units are connected j upstream of the Turbine Building exhaust

sampler, Radwaste Building exhaust sampler, and i Main Stack effluent monitor. Each monitor contains two redundant ion chamber detectors  ;

i and with a range of 0.1 to 10,000 R/hr.

Displays are mounted in the Main Control Room.

! In addition, three Control Room annuciators are l

associated with each HREMS monitoring unit -

FAILURE (loss of power or circuit malfunction),

l ALERT (high radiation) , and HIGH (high-high radiation). The ALERT and HIGH annuciators indicate when release levels begin to approach Site Emergency.or General Emergency Conditions (as defined by the FitzPatrick Site Emergency Plan). Dose rates are also recorded on three, two-pen recorders located adjacent to the instanteous rate displays. (Futher information on the HREMS can be found in Section 11.4.5.1 of the updated FitzPatrick Final Safety Analysis Report.)

7. Existing process and effluent monitors will also provide indication of releases or breaches l in systems used post-accident.

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ATTACHMENT 2 TO JPN-es5-91 New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 i

DPR-59 Plans and Schedules for the Implementation of Two

. Regulatory Guide 1.97 Revision 2 Variables -

l Standby-Liquid Control-Tank Level Instrumentation and Standby Liquid Control System Pump Flow l

In the Authority's June 28, 1985 letter (Reference 2),

the submittal of a schedule for two Regulatory Guide 1.97 variables (Standby Liquid Control Tank Level Instrumentation and Standby Liquid Control System Pump Flow) was deferred pending the development of plans and schedules to implement the ATWS (Anticipated Tranients l Without Scram) rule, 10 CFR 50.62. This attachment describes the Authority's plans for these two variables.

! 1. Standby Liquid Control Tank Level Instrumentation l Our plans and schedules for complying with the

! requirements of 10 CFR 50.62 (the ATWS rule) were l submitted to you in letter dated October 11, 1985 l (Reference 6). As stated in that letter, the Authority is i not altering the SLC system itself.

The equivalent SLC system flow requirement for i FitzPatrick is 87.3 gallons per minute of.a solution of thirteen weight-percent concentration of sodium pentaborate. This requirement will be met using an enriched boron solution. To achieve the required enrichment, enriched boric acid will be added to natural borax to produce double enriched sodium pentaborate

{ solution. The final solution will be obtained by mixing l the existing natural sodium pentaborate with a

} proportional amount of double enriched sodium pentaborate solution. The existing SLC system is capable of handling i the enriched solution with no changes to key SLC system process parameters (flow rates, discharge pressure, required-NPSH, etc.). I I

! Therefore, no changes to the existing SLC tank level instrumentation are considered necessary at this time to l . comply with the requirements of 10 CFR 50.62.

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2. Standby Liquid Control System Pump Flow Instrumentation i As stated in the Authority's November 30, 1984 Regulatory 4 Guide 1.97 implementation report . (Reference 1) , there is currently no direct indication of Standby Liquid Control System Pump Flow in the FitzPatrick Control Room.

However, the .SLC system pump discharge header pressure is

! displayed in the Control Room. This indication, along with several other indications available to the operator in the-control Room, provide reasonable assurance of proper SLC system operation. Specifically, six indications can be used to confirm proper operation:

I 1. : Loss of continuity .to squib valve annuciator.

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{ ?m Loss of amber " Squib-Valve-Ready" lights.

3. Illumination of SLC " Pump-Running" light.

4. SLC pump discharge header pressure greater than reactor pressure.

5. Decreasing SLC tank level.

l 6. Reactivity decrease-in reactor as measured by neutron flux monitoring.

. The Authority considers that these indications are adequate'to confirm proper SLC system operation; additional instrumentation would not significantly i increase the ability to detect system maloperation. The

{ BWR~ Owners Group has taken the same position in their i July 1982 report (Reference 12).

! Therefore, no changes to the existing SLC system instrumentation are necessary to' comply with either 10 CFR 50.62 (the ATWS Rule) or Regulatory Guide 1.97 l Revision 2. The instrumentation is acceptable as

! installed.

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ATTACHMENT 3 TO JPN-85-91 New York Power Authority James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 DPR-59 Comments on Preliminary Technical Evaluation Report (TER) regarding Conformance to Regulatory Guide 1.97 Revision 2 and Corrections to June 28, 1985 Letter Regarding Implemetation Schedule

1. The preliminary TER did not consider the new schedule information included with Reference 2. The revised schedules submitted with Reference 2 and the inforrention in this letter (and attachments) should be incorporated in the final report.

2. The Authority has noted an error in the cover letter of Reference 2. Specifically, the third paragraph on the first page should read:

"One new variable (Item 18) that was omitted from the prior schedule has been added. Installation schedules for one variable (Item 23) has been shortened one fuel cycle. Four variables (Items 1, 10, 21 and 22) have extended installation schedules."

The revised implementation schedule included with Reference 3 (Attachment 1) is correct - no changes are necessary.

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References for Attachments to JPN-85-91 i-

, 1. NYPA letter, J. C. Brons to D. B. Vassallo, dated November 30, 1984 (JPN-84-77) regarding response to Generic Letter 82-33.

2. NYPA letter, J. C. Brons to D. B. Vassallo, dated June 28, 1985 (JPN-85-53) regarding revised schedule for implementation of Regulatory Guide 1.97.

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3. NRC letter, D. B. Vassallo to J. C. Brons, dated

{ November 5, 1985 transmitted EG&G preliminary

Technical Evaluation Report regarding the same subject.

4. NRC Generic Letter 82-33 dated December 17, 1982 r

" Supplement 1 to NUREG-0737 - Requirements for

Emergency Response Capability," Section 6,

" Regulatory Guide 1.97 - Application to Emergency Response Facilities."

ll - 5. EG&G Preliminary Informal Report, EGG-EA-7040,

dated October 1985, "Conformance to Regulatory

Guide 1.97, James A. FitzPatrick Nuclear Power 1

Plant."

6. NYPA letter, J. C. Brons to D. B. Vassallo, dated

October 11, 1985 (JPN-85-73) regarding

! implementation of ATWS rule modifications.

, 7. PASNY letter, J. P. Bayne to D. B. Vassallo,

dated April 21, 1982 regarding post-TMI j requirements (Generic Letter 82-05).

8. PASNY letter, J. P. Bayne to T. A. Ippolito,

dated June 5, 1981 regarding post-accident

! shielding analysis. Includes shielding analysis

! results, source terms used, and areas considered.

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9. NRC letter, R. W. Starostecki to C. A. McNeill, 1

Jr. , dated April 26, 1983 regarding Inspection Report 50-333/83-07. Inspection conducted March 28-31, 1983 of actions taken to comply with the

! requirements described in NUREG-0737 Item II.B.2, i

" Design Review of Plant ~ Shielding."

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10. NRC letter, G. H. Smith to J. P. Bayne, dated i April 29 , 1982 regarding IEE Inspection j 50-333/82-03. Opened NRC Inspection Item Page 9

50-333/82-03-016 on Area Radiation Monitors i System upgrade (Appendix B, Item 11).

11. NRC letter, T. T. Martin to C. A. McNeill, Jr.,

dated November 16, 1983 regarding Inspection 50-333/83-22. Closed NRC inspection Item 50-333/82-03-016 regarding Area Radiation Monitoring System upgrade study.

12. BWR Owners' Group Report dated July 1982,

" Position on NRC Regulatory Guide 1.97 Revision

, 2."

13. James A. FitzPatrick Nuclear Power P3 ant updated Final Safety Analysis Report (FSAR) Section 3.9,

" Standby Liquid Control System."

14. NYPA letter, J. P. Bayne to D. B. Vassailo, dated May 20, 1983 (JPN-83-45). Attachment 1 to this

letter identifies equipment requiring 1

qualification in accordance with 10 CFR 50.49.

15. NRC letter, D. R. Muller to J. C. Brons, dated i

December 11, 1985 transmits order modifying i license to confirm additional license commitments l on the Safety Parameter Display System (SPDS),

Regulatory Guide 1.97 and Technical Support Center (TSC).

16. " Emergency Preparedness for What? (A Review of I the TMI-2 Incident)" by Andrew P. Muller, Safety and Environmental Protection Division, Brookhaven l National Laboratory (BN' 29257).

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