ML052060249

Multiple Letters and Post Exam Comments (Folder 1)

ML052060249
Person / Time
Site:	Peach Bottom, Limerick
Issue date:	07/11/2005
From:	Jason White Exelon Generation Co
To:	Caruso J Operations Branch I
Conte R
References
Download: ML052060249 (153)

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July 11,2005 Mr. John Caruso, Senior Operations Engineer U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406-1415

Subject:

2005 Limerick Generating Station Limited Senior reactor Operator Initial Examination Comments

Dear Mr. Camso,

Per NUREG-I 021, Rev. 9 Section ES-402.E, a post-Examination review was conducted of the 2005 LSRO Initial Written Examination administered on June 15, 2005 at Limerick Generating Station. As a result of this review, two questions are being submitted with comment for your consideration.

The questions, along with the justification for regrade and all applicable references are attached for your consideration.

Res ectfully, &he %D' sD5u B 'G, Director, Training - Limerick Generating Station Exe I o n IM Limerick Training Center 341 Longview Road Linfield, PA 19468 1041 Telephone 610 718 4000 Fax 610 718 4028 www exeloncorp corn Nuclear Exelon Nuclear Limerick Generating Station PO Box 2300 Sanatoga.

PA 19464-0920 June 22,2005 Mr. John Caruso, Senior Operations Engineer U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406-1415

Subject:

2005 Limerick Generating Station Limited Senior reactor Operator Initial Examination Comments

Dear Mr. Caruso,

Per NUREG-1021, Rev. 9 Section ES-402.E, a post-Examination review was conducted of the 2005 LSRO Initial Written Examination administered on June 15, 2005 at Limerick Generating Station. As a result of this review, two questions are being submitted with comment for your consideration.

The questions, along with the justification for regrade and all applicable references are attached for your consideration.

Respectfully Joseph L. White Director, Training - Limerick Generating Station /.

June 17,2005 Mr. John Caruso, Senior Operations Engineer U. S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406-1415

Subject:

2005 Limerick Generating Station Limited Senior Reactor Operator Initial Examination Comments

Dear Mr. Caruso:

Per NUREG-1021, Rev. 9 Section ES-402.E, a post-examination review was conducted of the 2005 LSRO Initial Written Examination administered on June 15, 2005 at Limerick Generating Station. As a result of this review, two questions are being submitted with comment for your consideration. The questions, along with the justification for regrade and all applicable references are attached for your consideration.

Respectfully, C. E. Rich u Manager, Operations Training - Limerick Generating Station Enclosures Question Number: 26 (Missed by all three candidates) Facility Regrade Request: Accept "d" as the correct answer Justification:

This question was modified from NRC Generic Fundamentals Examination Question Bank question 852. The question provides a reactor "shutdown for one week from long-term power operation and shutdown cooling in service."

It then provides that cooling water is lost to the shutdown cooling heat exchangers:

The candidate is then asked to determine which coefficient of reactivity will act first to change core reactivity and what the effect will be on Shutdown Margin. The given answer on the answer key provides that moderator temperature coefficient will be the first to act. The Facility Licensee agrees that moderator temperature coefficient will act first, since moderator temperature will rise as a direct result of the loss of cooling to the RHR heat exchangers.

The candidate must now decide if this will result in a decrease in Shutdown Margin (choice "a"), or an increase in shutdown margin (choice "d). For most of the core life, the reactor is considered to have a negative moderator temperature coefficient, where the effect of increasing moderator temperature will be to add negative reactivity to the core. This is due to the moderator density decreasing as a result of the temperature increase, causing neutrons to travel farther before slowing down to thermal energies, and having a higher probability of resonant absorption. Since more neutrons undergo resonant absorption, fewer neutrons are available for thermal fission, and the effect is to add negative reactivity to the core. This addition of negative reactivity moves the reactor farther from criticality, which increases Shutdown Margin.

This would make "d the correct answer. If the assumption is made that the core is at the end of life with low moderator temperature, the reactor could have a positive moderator temperature coefficient, which will result in the addition of positive reactivity as moderator temperature is increased.

This occurs because as moderator temperature rises, less moderator atoms are present in the core to compete with the fuel for the thermal neutrons.

This causes the thermal utilization factor to increase, resulting in more thermal neutrons available to cause fission in the fuel. The addition of positive reactivity moves the reactor closer to criticality, which decreases Shutdown Margin. This would make "a" the correct answer. Upon further investigation, information was obtained from LaSalle on reactivity effects of moderator temperature at various points in core life.

This information is not normally calculated for Limerick or Peach Bottom, but LaSalle is very similar as a C- lattice plant with 764 fuel bundles and 185 control rods. Core response at Limerick and Peach Bottom would therefore also behave in a similar fashion.

As can be seen in the attached spreadsheets for various times in core life, the moderator temperature coefficient can become positive as fuel exposure increases at low moderator temperatures.

This is common for BWR plants and can have operational impacts under these special conditions. However, under "all rod in" conditions, such as during an outage, the moderator temperature coefficient is alwavs negative.

This can be seen on the attached spreadsheets since the curve for the ARI condition never crosses the 0.000 reactivity point.

This is true for all exposure values calculated and for all temperatures. Based upon this data, answer "a" cannot be correct. Therefore, the Facility Licensee requests "d' be accepted as the correct answer. References Provided: . General Physics BW R Generic Fundamentals Reactor Theory Student Text, Chapter 2 (Neutron Life Cycle) . General Physics BW R Generic Fundamentals Reactor Theory Student Text, Chapter 4 (Reactivity Coefficients) . NRC Generic Fundamentals Examination Question Bank - BW R, Questions 852, 8948, 81248, 81752, B3652. . LaSalle spreadsheets of reactivity variations with moderator temperature at various times in core life (attached).

Question Number: 26 (Missed by all three candidates)

Facility Regrade Request: Accept "a" or "d" as the correct answer Justification:

This question was modified from NRC Generic Fundamentals Examination Question Bank question B52. The question provides a reactor "shutdown for one week from long-term power operation and shutdown cooling in service."

It then provides that cooling water is lost to the shutdown cooling heat exchangers.

The candidate is then asked to determine which coefficient of reactivity will act first to change core reactivity and what the effect will be on Shutdown Margin. The given answer on the answer key provides that moderator temperature coefficient will be the first to act. The Facility Licensee agrees that moderator temperature coefficient will act first, since moderator temperature will rise as a direct result of the loss of cooling to the RHR heat exchangers.

The candidate must now decide if this will result in a decrease in Shutdown Margin (choice "a"), or an increase in shutdown margin (choice "d"). For most of the core life, the reactor is considered to be in an "undermoderated" condition, where the effect of increasing moderator temperature will be to add negative reactivity to the core.

This is due to the moderator density decreasing as a result of the temperature increase, causing neutrons to travel farther before slowing down to thermal energies, and having a higher probability of resonant absorption.

Since more neutrons undergo resonant absorption, less neutrons are available for thermal fission, and the effect is to add negative reactivity to the core. This addition of negative reactivity moves the reactor further from criticality, which increases Shutdown Margin.

This would make "d" the correct answer. If the assumption is made that the core is at the end of life with low moderator temperature, the reactor could be in an "overmoderated" condition, which will result in the addition of positive reactivity as moderator temperature is increased. This occurs because as moderator temperature rises, less moderator atoms are present in the core to compete with the fuel for the thermal neutrons. This causes the thermal utilization factor to increase, resulting in more thermal neutrons available to cause fission in the fuel.

The addition of positive reactivity moves the reactor closer to criticality, which decreases Shutdown Margin. This would make "a" the correct answer. With the reactor in a shutdown condition, as given in the stem of the question, all control rods would be fully inserted. This will make the moderator-to-fuel ratio slightly smaller since some of the moderator is displaced by the control rods. This would have the effect of moving the core more toward an undermoderated condition, but still could be either undermoderated or overmoderated depending on core life and moderator temperature.

Again, either "a" or "d" could be considered a correct answer. Since the stem of the question does not provide the core age of the reactor, or whether it is "undermoderated" or "overmoderated", either "a" or "d" could be correct depending on the assumptions made by the candidate.

Therefore, the Facility Licensee requests

'a" or 'd" be accepted as a correct answer. References Provided:

General Physics BWR Generic Fundamentals Reactor Theory Student Text, Chapter 2 (Neutron Life Cycle) General Physics BWR Generic Fundamentals Reactor Theory Student Text, Chapter 4 (Reactivity Coefficients) . NRC Generic Fundamentals Examination Question Bank - BWR, Questions B52, B948, B1248, B1752,B3652.

LGSlPBAPS 2005 NRC LSRO Licensing Examination Question:

26 26 of 50 A nuclear reactor has been shutdown for one week from long-term power operation and shutdown cooling is in service. Upon a loss of cooling water to the shutdown cooling heat exchangers, which one of the following coefficients of reactivity will act first to change core reactivity and determine the effect on Shutdown Margin? (Assume continued forced circulation through the core) Coefficient to Act First Effect on Shutdown Margin

a. Moderator temperature coefficient Decrease b. Fuel temperature coefficient Increase

c. Fuel temperature coefficient Decrease

d. Moderator temperature coefficient Increase LGS/PBAPS 2005 NRC LSRO Licensinq Examination Cognitive (H, L) L PRA (Y/N) Unit (0, 1,2, 3) 0 N - -1 __~ - ~ ___ __ - --_ 7 Answer Key and Question Data SRO N Question # 26 Source: Reference( s): Learning Objective:

Knowledge/Ability:

__-- Choice I Basis or Justification Modified NRC QID:B52 BWR Fundamentals Chapter 2 BWR Fundamentals Chapter 2 Objective 9 292004 K1.14 1 Importance:

3.3 __-- a. 1 Correct Answer

b. I C. I I I I d. I Required Attachments or Reference Prepared by: CBG I BWR GENERIC FUNDAMENTALS REACTOR THEORY CHAPTER 4 REACTIVITY COEFFICIENTS If the nucleus mmained at standstill.

it would capture every neutron it came in contact with having an energy level of 2 I eV. 0 The nucleus is now vibrating in all directions due to the addition of heat energy (assume 5 eV). Thc nucleus will now capture all neutrons within a range of I6 eV to 26 eV. provided they "look like" 21 eV neutrons. The Nucleus is moving ._ - ~ --= - This neutron must "catch up" to the nucleus. In order to look like a 21 eV - this direction at 5 eV This neutron arrives LO--- - ----- head-on. To appear as a 21 eV neutron, it must be incoming at 16 eV. neutron. it must be incoming at 26 eV. 1.. .4 . This neutron must be incoming at an energy of 2 I eV. STUDENTTEXT

~ ~ C2000 General Physics Corporation.

Columbia Maryland All nghfs reserved Yo put of this book may be reproduced in MY fmm or by any mcani without pcrmisrion in wnring tiom Gmil Physics Corponoon TABLE OF CONTENTS .. TABLES AND FIGURES ............................................................................................................

II ... OBJECTIVES

..............................................................................................................................

111 KIA . OBJECTIVE CROSS REFERENCE

...............................................................................

iv REACTIVITY COEFFICIENTS

..............................................................................................

1 3 MODERATOR TEMPERATURE COEFFICIENT (a,,,) ............................................................ - Change in Moderator Temperature Coefficient with Core Agc ..............................................

4 Changes in Void Coefficient with Changes in Void Fraction ................................................

7 Changes in Void Coefficient with Changes in Fuel Temperaturc

..............................................

9 Changes in Void Coefficient with Changes in Core Age ...........................................................

9 VOID COEFFICIENT (a") ........................................................................................................

6 DOPPLER COEFFICIENT (aD) ................................................................................................

11 The Doppler Effect

...................................................................................................................

11 Self-shielding

...........................................................................................................................

14 Fuel Temperature Coefficient or Doppler Coefficient

.............................................................

18 Changes in Doppler Coefficient with Changes in Fuel Tempcraturc

......................................

20 Changes in Doppler Coefficient with Changes in Core Age ....................................................

20 Changes in Doppler Coefficient with Changes in Moderator Dcnsity ......................................

33 POWER COEFFICIENT (apowcr) ...............................................................................................

24 REACTIVITY DEFECTS

..........................................................................................................

26 REACTIVITY BALANCE AND DESIGN CONSIDERATIONS

...........................................

28 GLOSSARY ................................................................................................................................

31 BWR / REACTOR THEORY / CHAPTER 4 I 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 TABLES AND FIGURES Figure 4-1 Moderator Temperature and Density Changes .......................................................

3 Figure 4-2 vs . Moderator-to-Fuel Ratio

.............................................................................

3 Figure 4-3 Moderator Temperature Coefficient

.......................................................................

5 Figure 4-4 Core Void Formation

..............................................................................................

7 Figure 4-5 Flux and Void Content at Low Power ....................................................................

7 Figure 4-6 Flux and Void Content at High Power

...................................................................

8 Figure 4-8 U-238 Cross Section Curve ..................................................................................

12 Figure 4-9 Doppler Effect in Neutron Capture U-238 ...........................................................

12 Figure 4-1 1 U-238 Cross Section Curve ................................................................................

14 Figure 4- 13 U-238 Cross Section Curve ................................................................................

17 Figure 4-14 Fuel Temperature Gradients

...............................................................................

17 Figure 4-1 5 Fuel Temperature Effects on Self-shielding

......................................................

17 Figure 4-7 Void Coefficient

...................................................................................................

10 Figure 4-1 0 Doppler Effect ....................................................................................................

13 Figure 4-1 2 Self-shielding Effects

.........................................................................................

14 Figure 4-1 6 Doppler Coefficient of Reactivity

......................................................................

20 Figure 4- 17 Pu-240 Total Neutron Cross Section ..................................................................

21 Figure 4- 1 8 Doppler Coefficient of Reactivity

......................................................................

22 Figure 4-19 Doppler Defect ...................................................................................................

26 'T \ 1 ( BWR / REACTOR THEORY

/CHAPTER 4 11 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 OBJECTIVES Upon completion of this chapter, the student will be able to perform the following objectives at a minimum proficiency level of 8O%, unless otherwise stated, on an oral or written exam.

1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. Define and explain the moderator temperature coefficient of reactivity.

Describe the effects on the magnitude of the moderator temperature coefficient of reactivity from changes in moderator temperature and core age.

Explain resonance absorption.

Explain Doppler broadening and self-shielding.

Define and explain the fuel temperature (Doppler) coefficient of reactivity.

Describe the effects of core age, fuel temperature, core void fraction, and moderator temperature on the fuel temperature (Doppler) coefficient. Define and explain the void coefficient of reactivity.

Describe the effects of core age, fuel temperature, and core void fraction on the void coefficient of reactivity. Compare the relative magnitudes of the moderator temperature, Doppler, and void coefficients of reactivity.

Describe the components of the power coefficient.

Explain the differences between reactivity coefficients and reactivity defects.

Describe and explain the effect of power defect and Doppler defect on reactivity.

\ : a Q 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 ... BWR / REACTOR THEORY /CHAPTER 4 111 9 K/A - OBJECTIVE CROSS REFERENCE K1.03 K1.04 K1.05 REACTOR THEORY: 292004 REACTIVITY COEFFICIENT: Explain resonance absorption.

Explain Doppler broadening and self-shielding.

Define the Doppler coefficient of reactivity.

KIA # ~~ ~~ ~~ ~ 7- Describe the effect on the magnitude of the Doppler coefficient of reactivity for changes in the Fuel temperature.

Describe the effect on the magnitude of the Doppler coefficient of reactivity for changes in the Core age.

KIA STATEMENT K1.10 I Define the void coeficient of reactivity.

1 K1.O1 K1.02 Define the temperature coefficient of reactivity. Describe the effect on the magnitude of the temperature coefficient of reactivity from changes in moderator temperature and core age. K1.06 Describe the effect on the magnitude of the Doppler coefficient of reactivity for changes in the Moderator temperature.

K1.07 Describe the effect on the magnitude of the Doppler coefficient of reactivity for changes in the Core void fraction.

Kl .OS K1.09 1 IMPORTANCE RELATED OBJECTIVE RO SRO NUMBERS 1 2.6 2.7 4 7 3* -.- -. ? I* 6 3.2 I 3.2 1 7 ~ ~~~ BWR REACTOR THEORY CHAPTER 4 / REACTIVITY COEFFICIENTS iv 0 2000 GENERAL PHYSICS CORPORATION REV 3 I WA - OBJECTIVE CROSS REFERENCE K1.11 \ Describe the effect on the magnitude of void coefficient from changes in the core void fraction.

I REACTOR THEORY:

292004 REACTIVITY COEFFICIENTS 2.5 KIA # 2.6 KIA STATEMENT K1.12 IMPORTANCE RO SRO Describe the effect on the magnitude of void coefficient from changes in the fuel temperature.

2.2* 2.3* K1.13 Describe the effect on the magnitude of void coefficient from changes in the core age. K1.14 Compare the relative magnitudes of the temperature, Doppler, and void coefficients of reactivity.

2.1* I 2-2* RELATED OBJECTIVE NUMBERS 8 8 8 9 The following objectives, while not cross-referenced to specific UAs, ensure mastery of findamental concepts:

IO, 1 1, and 12. Note: Importance ratings that are marked with an asterisk (*) or question mark (?) indicate variability in rating responses by reviewers.

An asterisk (*) indicates that the rating spread was very broad.

An asterisk (*) can also indicate that more than 15% of the raters felt the knowledge or ability is not required for the RO/SRO position at their plant. A question mark (?) indicates that more than 15% of the raters felt that they were not familiar with the knowledge or ability as related to the particular system or design feature. A dagger (t) indicates that more than 20% of the raters indicated that the level of knowledge or ability required by an SRO is different from the level of knowledge or ability required by an RO. BWR / REACTOR THEORY

/CHAPTER 4 V 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 THIS PAGE INTENTIONALLY BLANK . .... / BWR / REACTOR THEORY /CHAPTER 4 vi 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 I REA CTl VI TY r I COEFFICIENTS If the parameter (x) increases and positive reactivity is added, then a, is positive.

If the parameter increases and negative reactivity is added, then a, is negative. Recall from Chapter 2 that reactivity (p) is defined as the fractional change in neutron population per generation.

and that the value is determined by the formula p =Am. In Chapter 3. several examples were presented that added positive and negative reactivity to a reactor. The examples showed the effect on reactor power and how various values of reactivity affect the rate at which reactor power changes. It is important for a reactor operator to know how a change in any of the plant parameters affects reactor power.

This knowledge allows the operator to predict reactor response during plant evolutions and transients involving parameter changes.

Chapter 4 explains how changes in specific core operating parameters change the six factors of and, therefore, change reactivity (Ap) and reactor power.

The core operating parameters are moderator temperature, he1 temperature, and core steam void fraction. The change in reactivity (Ap) due to the per unit change in the associated parameter (Ax) is called the reactiviq coeflcient (a) for that parameter (x). In general terms, a reactivity coefficient is defined as: r AP a, =- Ax Where: a, = reactivity coefficient for plant parameter x Ap = change in reactivity (M) AX = changeinsomeplant parameter Equation 4-1 BWR / REACTOR THEORY / CHAPTER 4 1 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3

.. MODERATOR TEMPERATURE COEFFICIENT (a,) The moderator temperature coeficient predicts changes in reactivity resulting from changes in moderator temperature.

It is defined as the change in reactivity per unit change in the temperature

("F) of the moderator:

Where: am = moderator temperature coefficient (MTC)(Ak/k/"F)

Ap = change in reactivity (Akk) AT,,,^^ = change in moderator temperature (OF) Equation 4-2 The symbols aT or MTC are also used to represent the moderator temperature coeficient.

The symbol a,,, will be used in this text. A reactor operating at 530°F has a I k,r= 1.000. The moderator temperature is increased to 540°F and k,lr decreases to 0.999. Calculate the value of the moderator temperature coefficient.

Solution:

Example 4-1 A good approximation for the average value of a,,, is -1 x 1O4AkM0F for the normal range of moderator temperature at power.

For the moderator (water), a temperature increase results in a density decrease.

As shown in Figure 4-1, the magnitude of the density change for a given temperature change gets larger with increasing temperatures.

\ .II .. BWR / REACTOR THEORY /CHAPTER 4 2 of 39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3

- a t 04" I W c r' I' I I I - 7 MODERATOR TEMPERATURE I II It II II 1 - - - - - - - - ---a II 11 AT AT MODERATOR TEMPERATURE Figure 4-1 Moderator Temperature and Density Changes This results in the magnitude of the moderator temperature coefficient being larger (more negative) at higher temperatures. The moderator temperature coefficient for a one degree change at a high temperature (499 to 500OF) is more negative than the moderator temperature coefficient at a low temperature (99 to 100°F). Since reactivity is defined in terms of the t effective multiplication factor (h) it is necessary to examine how moderator temperature changes affect the effective multiplication factor or the six factors. Recall:

ken = JfP 4h f rl Equation 4-3 We have shown that an increase in moderator temperature results in a decrease in water density. This causes an accompanying increase in slowing down and thermal diffision lengths because the moderator atoms are farther apart, requiring neutrons to travel farther between collisions.

Increasing the slowing down length increases the probability that a neutron can reach the fuel while still at resonance energy. Since the slowing down length increases, the slowing down time also increases.

Thus, neutrons spend more time at resonance energy levels. Reducing f the probability of a neutron escaping resonance capture decreases the resonance escape probability (p). The plot for p shows this effect in Figure 4-2. UNDER I OVER MODERATED ct-) MODERATED Figure 4-2 kclg vs. Moderator-to-Fuel Ratio A decrease in the moderator density also causes the thermal neutron absorption in the moderator to decrease due to fewer moderator atoms in the core area. This increases the probability of thermal neutron absorption in the fuel. In addition, the thermal utilization factor (f) slightly increases (Figure 4-2). Recall from Chapter 2 the equation:

P tile1 Equation 44 This can be rewritten as: ,fuel Equation 4-5 As the temperature increases, the concentration of moderator atoms (Nmd) decreases; therefore, the thermal utilization factor increases.

BWR / REACTOR THEORY / CHAPTER 4 3 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 Decreasing moderator density increases the migration length of the neutrons, which increases the fraction of neutrons leaking out of the core.

and therefore decreases the nonleakage factors. For large commercial power reactors. neutron leakage is insignificant.

The fast fission factor increases slightly due to increased slowing down length, but the effect is minimal. Because the reproduction factor is not dependent on moderator density, it does not change as moderator temperature changes. As Figure 4-2 shows, moderator temperature changes result in essentially two competing processes:

the resonance escape probability (p) and the thermal utilization factor (0. The resonance escape probability has the dominant effect, causing kff and reactor power to decrease as moderator temperature increases.

Since increasing moderator temperature (decreasing the moderator-to-fuel ratio) decreases kern, the moderator temperature coefficient is negative. The region to the left of the maximum effective neutron multiplication factor is the undermoderated region. Note that in this region an increase in temperature causes a reduction of the effective neutron multiplication factor. This results in a negative moderator temperature coefficient. Operating in the undermoderated region is very important in terms of reactor control. If reactor power suddenly increases, the moderator temperature wifl rise, inserting negative reactivity into the system and thus limiting the power excursion. Commercial reactors are designed with a moderator-to-fuel ratio such that the moderator temperature coefficient is negative.

The region to the right of the maximum effective neutron multiplication factor is the overmoderated region. In the overmoderated region, the reduction in moderator density has a greater effect on the thermal utilization factor than the resonance escape probability.

The increased thermal utilization causes a positive, reactivity addition with increasing moderator temperature.

If the reactor were allowed to operate on the overmoderated side of the curve. any increase in power would cause an increase in moderator temperature. adding positive reactivity and accelerating the power increase. - At higher temperatures, the moderator temperature coefficient becomes more negative due to a larger change in density for the same change in temperature.

In a BWR, the moderator is at saturated conditions once normal operating temperature is reached, and the moderator temperature does not change significantly afterwards.

Thus, the moderator temperature Coefficient affects reactor power more during heatups and cooldowns.

CHANGE IN MODERATOR TEMPERATURE COEFFICIENT WITH CORE AGE As the core ages, fuel density decreases and, in order to maintain power, control rods are withdrawn from the core.

Both the moderator-to-fuel ratio and effective core size increase. Effective core size also can be discussed in terms of control rod density. Control rod density, in a BWR, is the ratio to control rod notches inserted into the core to the total number of control rod notches available in the core. Thus, 100% rod density means all rods are fully inserted and effective core size is 0%. At 0% rod density, all rods would be fully withdrawn and the effective core size would be at maximum. A control rod density of 25% means that the effective core size is 75%. BWR / REACTOR THEORY / CHAPTER 4 4 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 i L In sum, the smaller the rod density the larger the effective core size. The lower rod density decreases the amount of neutron absorbers in the core. This minimizes the likelihood of a neutron being absorbed in a control rod because it travels farther when moderator temperature increases.

Therefore.

the thermal utilization factor (f) increases as the core ages; this is the dominant effect.

Both fast (If) and thermal

(&) nonleakage probabilities increase slightly as the effective core size increases with core age. Counteracting these increases is the effect of resonance escape probability (p). As the core ages, Pu-240 builds up. This increases the chance of resonance absorption, which decreases the resonance escape probability. Figure 4-3 plots am versus temperature for both the beginning of cycle (BOC) and the end of cycle (EOC). The moderator temperature coefficient (a,,,) is negative at BOC and becomes more negative at higher moderator temperatures.

As the core ages, am becomes less negative and slightly positive at very low temperatures near the EOC cycle. By reducing the control rod density at criticality, the moderator temperature coefficient (a,,,) becomes less negative for low temperature-zero power conditions (i.e., criticality occurs with more control rods withdrawn).

AVERAGE TEMPERATURE

('F) Figure 4-3 Moderator Temperature Coeflcient The potential for the occurrences of positive am at the end of cycle has become larger as cycle lengths are increased from 18 to 24 months. A positive value of a, can be observed as a reactor period that becomes slightly shorter without additional operator action. The moderator temperature coefficient (a,) is negative by design, since all light water reactors in the U.S. are designed to be undermoderated for all normal operating conditions.

An average value of am is given as: < P Ak/k "F a,,, = Equation 4-6 d BWR / REACTOR THEORY / CHAPTER 4 5 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 I VOID COEFFICIENT (a") I In a BWR, the formation of steam bubbles has essentially the same effect (but greater in magnitude) on average moderator density and the moderator-to-fuel ratio as an increase in moderator temperature.

The core void fraction is the ratio of the volumetric fraction of steam in the core to the volume of steam plus liquid in the core: Volume of Steam Volume of (Steam + Liquid) Void Fraction = Equation 4-7 This fraction can be expressed as a percent, and is called percent voids.

The mechanism by which void content affects neutron flux is essentially the same as for moderator temperature changes.

As steam bubbles or voids are formed, liquid moderator is displaced and moderator density decreases. Since the voids look like large holes in the moderator to the neutrons, the effect of voids on hm is much greater than that caused by moderator heating. The definition of the void coefficient (a") is the change in reactivity per unit change in the overall core void fraction.

At) - - Pfina~ - Pinitid a, = A%voids %voids,,, - %voidsinitid Where: aV = void coefficient (W% voids) Ap = change in reactivity (Akk) % voids = void fraction expressed in % Equation 4-8 A reactor has an average core void fraction of 20% with kern = 1.000. Calculate the void coefficient if the void fraction is increased to 22% resulting in a kern decrease to 0.998. Example 4-2 A good approximation for the void coefficient is -1 x 1 Oe3 Ak/k/% void. The mechanisms that cause an addition of negative reactivity as void fraction increases are essentially the same as the mechanism affecting the moderator temperature coefficient. Increasing the core void fraction decreases moderator density and decreases the moderator-to-fuel ratio. Neutron leakage from the core increases, neutron absorption by moderator molecules decreases, and resonance absorption increases. The difference between the moderator temperature coefficient and steam void coefficient is that voids cause a much larger decrease in the moderator-to-fuel ratio.

As a result, the core steam void coefficient is larger in magnitude (more negative) than the moderator temperature coefficient.

BWR / REACTOR THEORY

/CHAPTER 4 6 of 39 Q 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 t CHANGES IN VOID COEFFICIENT WITH CHANGES IN VOID FRACTION To understand how the magnitude of the void coefficient changes with a change in void fraction, we must determine where voids are formed in the core and their relationship to the flux concentration in the core. Figure 4-4 represents a localized.

high power region within the core, showing two fuel pins and the void formation that occurs between these pins. FUEL PIN A t d m I) .- V FUEL PIN FLOW . -5 + 14 c) 3 Water (moderator)

IS at and voids start moving up the core Voa start to folm atthedadsurfaceand co(lepQc when reaching subcookd water sbowcontml rod FLOW bp , In hqh powcw fuel. the voids mtnlnne to fonn an annubar space Ween tho plm with a thin layer dwateratthepmsurfacs , Voida mkne wth each other to form even hger volds. ~~modarator den* em mon In the high power regions of the core. the area near the exit of the fuel channel may consist of only a thin layer of the moderator at the surface of the fuel with the majority of the area voiding between the fuel pins. At low power levels, the peak neutron flux is located in the upper one-half of the core.

Figure45 graphs the flux distribution and location of the voids in a low power reactor.

As shown, the voids mostly concentrate in the core's upper portion and away from the core's higher flux regions. TOP OF CORE NEUTRON FLUX CONTENT BOTTOM OF CORE Flux __* .X Void Conte_nt Figure 4-5 Flux and Void Content at Low Power Figure 4-4 Core Void Formation Starting above the tip of the withdrawn control rods, the moderator is heated to saturated conditions where voids begin to form. Initially, the subcooled water sweeps these voids away and they collapse.

As the moderator continues up the fuel channel, it reaches saturation temperature and the voids begin moving up the channel. The formation of voids increases as the fuel continues to add heat to the moderator, and the concentration of voids increases as the moderator flows upward. Also, as the voids move up the fuel channel, they begin to combine to form larger voids.

1 ..A BWR / REACTOR THEORY

/ CHAPTER 4 7 of 39 6 2000 GENERAL PHYSICS CORPORATION

,' REACTIVITY COEFFICIENTS REV 3 Figure4-6 graphs the flux distribution and In summary, as the core void fraction increases, location of voids for a high power reactor. The the fraction of the core moderated by water void fraction is still the greatest at the top of the decreases. Negative reactivity is added because core, however a much higher concentration is of the poor moderating ability of water vapor. distributed down into the core's higher flux Furthermore, at higher void fractions. a larger regions. amount of negative reactivity is added for the same increase in void fraction. This is the result of the greater fraction of voids in the higher flux region of the core, as shown in Figure4-4, Area 4. We can also show this concept with the - J TOPOFCORE BOrrOMOFCORE flux % Void content d _* Figure 4-6 Flu and Void Content at High Power This occurs because the control rods must be moved farther out of the core to increase reactor power level.

As the rods are moved toward the bottom of the core, the core average neutron flux increases and the flux profile tends to follow the control rods. In addition, the voids that are formed near the top of the rod tips are pulled farther down into the core and into higher neutron flux.

It is actually possible to see a net reactor power decrease with a control rod withdrawal. This will occur with control rods that are nearly fully withdrawn from the core. Withdrawing these control rods results in a local power increase at the rod tips. The local power increase is in a relatively low neutron flux region of the core, which then adds voids to the affected fuel channels. These voids travel up the core into higher neutron flux regions resulting in the addition of more negative reactivity from the void coefficient than the positive reactivity added by the rod movement. following example.

If at 10% voids, 90% of the core is moderated by liquid water, then a 1% increase in void fraction will void about 1/90 or 1.1% of the core. If the core is 30% voided, 70% of the core is moderated by water. Then a 1% increase in void fraction will void about 1/70 or 1.4% of the core. Thus, a 1% increase in voids at 30% void fraction will add more negative reactivity than a 1 YO increase at 10% void fraction.

Keep in mind that: Volume of Steam - Void Fraction = Volume of (Steam + Liquid) Equation 4-9 Hence, a 1% increase at a 10% void fraction will void about 1/90, or about 1 .I 1%, of the core. Typical core void fractions for most of the large commercial BWRs range from about 38% to 42% when at 100% reactor power.

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/ REACTIVITY COEFFICIENTS REV 3 CHANGES IN VOID COEFFICIENT ( WITH CHANGES IN FUEL -. TEMPERATURE

~ ~~ ~~ As fuel temperature increases, more neutrons undergo resonant capture; therefore, the resonance escape probability (p) decreases.

The fraction of neutrons that undergo resonant capture is dependent upon the width of the resonance peaks and the fraction of neutrons in the resonant energy spectrum.

At high fuel temperatures, the resonance peaks are relatively wide. The increase in the width of the resonance peaks results in a larger fraction of the core neutrons available for resonance capture. An increase in void fraction increases the slowing down length and slowing down time, which results in a larger fraction of the neutrons reaching the fuel at resonance energy. Therefore, a 1% increase in void fraction at high fuel temperature creates a greater decrease in the resonance escape probability than the same increase at low fuel temperatures. This results in a void coefficient of larger magnitude (more negative) at high fuel temperatures. The section covering the Doppler coefficient discusses the resonance capture in more detail.

L CHANGES IN VOID COEFFICIENT WITH CHANGES IN CORE AGE deplete more rapidl) than the fuel and control rods must be inscrted to hold power constant (recall discussion on kC,Cc',,).

At approximately one-third to one-hall' of the cycle. control rod density reaches it3 maximum (about 15% to 16% with the rexior at 100% power). Effective core size is :it 114 smallest value, and the rnagnitudc of' ttic \oiJ coenicient is at its maximum. 1 1114 point is reached when fuel depletion and hmi.ible poison depletion are the same. Opcratioii h*! ond this point requires control rods to h- \{ irlidrawn to maintain IOO% reactor PO\\ er The smaller CIIL*L~I~C core size means that the power and ttic \cuJ\ arc' produced in a smaller portion 01' thc c'orc There are more steam bubbles (\oidh t in the power-producing (high neutron flux) pmion3 of' the core, decreasing the moderator-to-fuel ratio and thus making that part of thc core niorc undermoderated. As control rod5 arc uiihJra\vn during the balance of the cycle. c'tlc'cti\c core size increases and the voids arc dihprscd over a larger area of the core. making thc effective core less undermodcratcd . To better understand how effective core size affects the magnitude of the void coefficient.

consider thc follo\\ing:

A 40% void fraction means that 4090 of the physical core volume is occupied bj steam. It is not possible to make a specific statement on how the void coefficient varies with core age.

However, it is generally true that the value of the void coefficient changes proportionally with control rod density or inversely proportional to effective core size. At the beginning of a fuel cycle, the control rod density is approximately 10% to 12% when the reactor is at 100% power, equilibrium conditions.

As the reactor operates during the early part of the cycle, the burnable poisons BWR / REACTOR THEORY

/ CHAPTER 4 9 of 39 0 2000 GENERAL PHYSICS CORPORATION L / REACTIVITY COEFFICIENTS REV 3 The volume of the 40%voids never changes. but effective core size does change, and this volume of steam voids is always located within the effective core size area. If we assume a (very exaggerated) effective core size of 1/2 of physical core size (50% control rod density) and the 40% void fraction is located entirely in this area of the core, then to the effective core the void fraction would actually look like 80% (80% effective void fraction) with a value of -1 x Ak/k for each 1%. Adding a 5% void fraction to the core at this time would look like a 10% void fiaction addition to the effective core, with a total reactivity value of (-1 x x 10=-1 x 10-2Ak/k.

Increasing the effective core size to 80% would result in an effective void fraction of 50% (40% voids/80% effective core size = 50% effective void fraction). Adding a 5% void fraction to an effective core size of 80% would look like a 6.25% void fraction addition to the effective core, With a total value of 6.25 x 1 0-3 Ak/k. From the above example, it is evident that the smaller the effective core size (the greater the rod density), the larger the magnitude of the void coeficient.

Figure 4-7 helps convey this discussion on steam void coefficient. The void coefficient (av) becomes more negative as core voit fraction increases because the core is more

• undermoderated.

0 10 20 30 40 50 60 70 80 CORE AVERAGE VOIDS (%) Figure 4-7 Void Coefjcimt The void coefficient (av) is always negative throughout core life because the core is always undermoderated. As the core ages, the effective core size initially decreases due to burnable poisons being depleted faster than the fuel, and aV becomes more negative with decreasing effective core size. As fuel depletes and core size increases due to control rods being withdrawn, av becomes less negative.

An average value for av is: - c Aklk %voids a, = -1 x io-' Equation 4-10 _- .I 4 BWR / REACTOR THEORY / CHAPTER 4 IO of 39 6 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 DOPPLER COEFFICIENT I (aD) I THE DOPPLER EFFECT The Doppler effect is defined as the apparent change in frequency of sound, light, or radio waves (or any other form of energy) caused by motion. The Doppler concept has been applied in various ways; the most familiar is radar. A radar transceiver puts out pulses of radio waves and then receives the waves that bounce off from incident objects to determine such things as size, shape, density, or speed, depending upon what it is designed to sense. It does this by measuring the shifts in frequency between the transmitted wave and what is received and then comparing against known facts programmed into the unit. An example of the Doppler effect is the change in pitch of a horn or other sound from a vehicle speeding towards and then away from us. We hear a higher pitch sound moving toward us because the sound waves are moving in our direction at the speed of sound with the speed of the vehicle added to it. Thus, the compressed waves result in a higher pitch sound.

As the same vehicle moves by and away from us, we hear the sound go from a high to a lower pitch. The lower pitch is the result of the speed of the sound waves coming at us, minus the speed of the vehicle moving away from us. This makes the sound waves seem longer, giving the lower pitch sound. The sound that is heard by the persons in the vehicle is heard only at a single true pitch, since they are moving with the sound.

c The term Doppler coefficient (aD) is also referred to as the fuel temperature coefficient because its value is primarily a function of fuel temperature.

It is more often called the Doppler coeficient because of the phenomenon known as the Doppler effect that occurs with the heating of the fuel.

Therefore, in order to adequately explain and discuss the Doppler coefficient, it is first necessary to study the relationship between fuel temperature within the reactor and the probability of resonance absorption (resonance capture) of neutrons.

As previously discussed in Chapter 1, neutrons give up energy in step changes through collisions with nuclei. Fast neutrons must pass through intermediate (epithermal) energies prior to reaching thermal energies.

All neutrons at epithermal energies have a probability of being lost due to resonance absorption.

The microscopic cross section for absorption (a,) for U-238 is 5.500 barns for neutrons at an energy level of 21 eV, but only 15 to 20 barns for a neutron with energy levels of 20 or 22 eV. Unfortunately for neutrons, U-238 has several other energy ranges which will resonantly capture neutrons, as shown in Figure4-8.

To make matters worse, U-238 is not the only resonant absorber in the reactor.

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/ REACTIVITY COEFFICIENTS REV 3 c 10,000 L 68°F I 1 TEMPERATURI I I ~ -. 1 4'500 I...... ELEVATED .. I1 I I1 500 1.00( NEUTRON ENERGY (eV) Figure 4-8 U-238 Cross Section Curve Figure4-8 shows the U-238 cross section for absorption as a fbnction of neutron energy for two different fuel temperature conditions.

The resonance peaks, shown by the solid cross section curve, assume that the target nucleus (U-238) is at a nominal ambient room temperature of 68°F (= 21°C) and the incident neutron provides the majority of all the kinetic energy in this neutron interaction with the U-238 target nucleus. In the reactor, however, this is rarely the case because the nuclear fuel would be generally at some elevated temperature.

This is either due to heatup of the reactor coolant or to power operation of the reactor.

Even atoms in crystals that are rigidly bound vibrate in their crystal lattice (if temperature is above absolute zero). As temperature increases, atom kinetic energy increases.

The vibration of these target nuclei results in a change of absorption cross section characteristics.

To demonstrate this phenomenon.

consider the three neutron-nuclear reactions depicted in Figure 4-9a, b, and c. respectively.

Suppose an incident neutron of 21 eV of kinetic energy impinges on a target nucleus at room temperature (roughly 0.025 eV), as shown in Figure4-9a.

By using the cross section graph for U-238 as given in Figure4-8, one would determine the absorption cross section for this event to be about 5.500 barns. A RELATIVE ENERGY 2: 21eV m .---., 21 eV '~ NUCLEUS NEUTRON 0.025 eV am x 5.500 barns n, i INCIDENT '*- -1 J a: 21 eV INCIDENT (RESONANCE)

NEUTRON ua 5,500 barns b: 20 eV INCIDENT (OFF-RESONANCE) NEUTRON RELATIVE ENERGY I: 21eV F - - 22 eV ,/LGh lev n J '#?9- '\./ a. 2: 5,500 barns c: 22 eV INCIDENT (OFF-RESONANCE)

NEUTRON Figure 4-9 Doppler Effect in Neutron Capture U-238 d BWR / REACTOR THEORY / CHAPTER 4 12 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 In Figure4-9b.

a 20eV neutron impinges on a U-338 nucleus that is vibrating toward it with a kinetic energy of 1 eV. The relative energy between the incident neutron and the target U-238 nucleus is also 21 eV. Hence, the resultant absorption cross section for this case also would be about 5.500 barns. In Figure 4-9c, the incident neutron possesses a kinetic energy of 22 eV and the target U-238 nucleus is vibrating away from it with a kinetic energy of 1 eV. The relative energy between the incident neutron and the target U-238 nucleus is, once again, 21 eV. Hence, the resultant absorption cross section for this case as well would be about 5,500 barns. Therefore, Figure 4-9a, b, and c depict the Doppler effect in neutron physics. As fuel temperature increases, the kinetic energy of the fuel atoms increases.

Hence, neutrons of even higher and lower kinetic energy have an increased probability of resonance absorption.

Figure 4-10 illustrates the Doppler effect as it relates to the relative motion (energy) between the neutrons and a U-238 nucleus, concentrating on the effect of the 21 eV resonance peak.

This example illustrates the affect of more energy applied to the nucleus.

If the nucleus remained at standstill.

it would capture ebery neutron it came in contact ~ith having an energy lebel of 1 I eV 0 The nucleus is now vibrating in all directions due to the addition of heat energy (assume 5 eV) The nucleus will no\* capurc all neutrons nithin a range of 16 eV to 26 eV. pm\ ided they "look like" 11 eV neutrons The Nucleus is mo\ ing this direction at 5 eV 0 This neutron arrives This neutron must "catch up" to the nucleus In order to looh like a 21 eV neutron. it must be incoming at 26 eV head-on To appear as a Z I eV neutron. it must be incoming at 16 rV This neutron must be incoming at an energy of 2 I eV Figure 4-10 Doppler Eflect When adding 5 eV of heat energy to the nucleus, it rapidly vibrates in all directions.

The nucleus still prefers a 21 eV neutron and only captures those neutrons that it "sees" as 21 eV neutrons.

Because of the relative motion between the nucleus and the surrounding neutrons, however, it now absorbs any neutron within a kinetic energy range of 16 eV to 26 eV, depending upon the angle that they approach the nucleus. The only criteria is that the neutrons appear as a 21 eV neutron to the nucleus upon arrival.

By adding more heat to the nucleus, its speed and area of vibrational motion increase.

However, because it is vibrating faster, it now spends less time at any given energy within its kinetic energy range. Put simply, the vibrating nucleus now has the capability of also capturing the "off-resonance" neutrons of 16 eV and 26 eV, respectively. The vibration of the nucleus reduces the probability for capturing a 21 eV "resonance" neutron, but the U02 fie1 pellet still captures it.

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.. Therefore, heating the nuclear fuel serves to "broaden and flatten" the U-238 resonance peaks on the absorption cross section curve because of the expanded range of neutron energies made available for capture, but also the nucleus spends less time at any given energy.

The dashed lines show this effect in Figure 4-1 I. This shift (widening and flattening) of the resonance peaks is called Doppler broadening. - WF 1.500 ...... ELEVATED TEMPERATURE I I II I I 500 l.0oa NEUTRON ENERGY (ev) Figure 4-1 1 U-238 Cross Section Curve Because the area under both the original and the broadened curve are theoretically the same, one assumes that the overall capture of neutrons by U-238 should not change significantly. However, within the reactor, the heating of the fuel and the broadening of the U-238 resonance peaks increases the resonant neutron absorption in the UO2 fuel pellets. To understand this important phenomenon, it is necessary to examine the effects of self- shielding that occurs within the fuel pellets. SELF-SHIELDING 3 Up to this point, the Doppler effect has not had an effect on the operational characteristics of the reactor. If all reactors were homogeneous.

the Doppler effect would not affect the reactor at all; however. commercial boiling water reactors are not homogeneous. The fuel is comprised of ceramic pellets that are housed in a helium gas-filled. zircaloy clad, cylindrical fuel pin. The neutrons are slowed down in the surrounding moderator. High energy neutrons pass through the fuel pellets and clad to the moderator.

The moderator slows down the neutrons into the epithermal and thermal ranges.

At low fuel temperatures, a neutron entering a fuel pellet with the exact resonant energy has a very high probability of absorption and will be most likely absorbed in the outer edge of the fuel pellet. Epithermal neutrons, other than resonant energies, are more likely to pass directly through the pellet without being absorbed. The outer fuel atoms tend to shield the inner fuel atoms from the resonant energy neutrons. The term for this is self-shielding

-' effect. > To describe self-shielding, consider a UOZ fuel pellet at room temperature and another one at operating fuel temperature in the reactor, as shown in Figure 4-12a and

b. 22 eV 20 eV 22 eV 20 eV ,a - m <_ R - n . - !!! !! 21 eV 21 eV a UO2 FUEL PELLET AT b. UG FUEL PELLET AT ROOM TEMPERATURE OPERATING REACTOR TEMPERATURE AT POWER Figure 4-12 Serfsirielding Effects . --d BWR / REACTOR THEORY

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/ REACTIVITY COEFFICIENTS REV 3 In Figure 4-12a. only resonance neutrons are captured as shown by the 21 eV resonance neutron. Off-resonance neutrons pass right through and are not "seen by the UOz fuel pellet. Because only the resonance neutron is captured upon entering the fuel pellet and the off-resonance neutrons are not captured, the inner region of the pellet is "self-shielded" by the outer periphery.

r Figure 4-12b depicts the U02 fuel pellet at an elevated temperature as experienced at power conditions. Due to the high vibration of the U-238 nuclei, both resonance and off-resonance neutrons are captured under these higher fuel temperature conditions, as discussed in the previous section on the Doppler effect.

Figure 4- 12b shows a reduction in self-shielding under these circumstances.

That is, the fuel pellet's central portion captures both off-resonance and resonance neutrons.

The following discussion addresses the concept of self-shielding from a calculational point of view. This serves to supplement the qualitative, c physical description of self-shielding just presented.

Two issues need considering to determine the amount of self-shielding within a fuel pellet. Both are primarily a function of fuel design and, although they are addressed as separate issues in this text, the second issue is an extension of the first issue. Combining the two issues determines the effect of fuel temperature on the neutron population in the core. The first issue relates to the physical size of the fuel pellets and the average distance that a neutron travels into the pellets prior to resonance absorption. Recall that the mean free path (A) is defined as the average distance that a neutron travels before being absorbed.

The atomic density (N) is typically 2 x 1 02' atoms/cm3 for U-238 in a fuel pellet. In this discussion, assume that in three mean free paths every neutron is absorbed.

1 h, =- No, Where: h, = mean free path (cm) N = atomic density (atoms/cm3) o8 = microscopic cross section for absorption (barns)

Equation 4-1 I If 100 neutrons, all at an energy level of 21 eV, enter a fuel pellet, all the neutrons are absorbed if the fuel pellet is three mean free paths wide.

At 21 eV, U-238 has a resonance peak of 5,500 barns. Calculate the mean free path of the 21 eV neutron and the value of three mean free paths. Since the average fuel pellet is 1.0 cm in diameter, all 100 neutrons at 2 1 eV entering the fuel pellet are absorbed.

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/ REACTIVITY COEFFICIENTS REV 3 For the neutrons not at the energy level of the resonance peak, say 22eV. the microscopic cross section is about 15 barns. This leads to a mean free path of 3.3 cm for these neutrons.

In order for all these neutrons to be absorbed in the U-238, the fuel pellet has to be about 1Ocm. or 4.0 inches in diameter. Therefore, in the 1 .O cm fuel pellet, very few off-resonance neutrons are absorbed.

If there are 100 neutrons at 22 eV, two of these off-resonance neutrons are absorbed. Thus, of the 200 total neutrons that entered the fuel pellet at the two energy levels, 102 neutrons are absorbed in the fuel pellet. As the fuel temperature increases, the microscopic cross section for the neutrons at the energy level corresponding to the resonance peak decreases, but increases for the energy levels around the resonance peak. The curve changes shape, but the area under the curve remains constant.

So for the 1.0 cm fuel pellet, 102 neutrons are still absorbed. However, not all of the neutrons at the energy level corresponding to the resonance peak are absorbed.

For comparison, assume that at 600"F, 99 resonant neutrons are absorbed and three off-resonance neutrons are absorbed, totaling 102 neutrons. Remember that the microscopic cross section has decreased for the 21 eV neutron and increased for the 22 eV neutron.

Therefore, a slight possibility exists that some neutrons at 21 eV will escape. Decreasing the microscopic cross section has the effect of decreasing self- shielding.

A 21 eV neutron is likely to travel farther into the fuel pellet prior to capture, and some may pass completely through the pellet. The off-resonance neutrons that normally would have passed completely through the pellet now have an increased probability of being captured within the pellet. We stated that the average fuel pellet has a diameter larger than the three mean free paths needed for complete neutron absorption.

Ir. other words. part of the fuel pin does not see a neutron flux at low fuel temperature with energy levels at the resonance peak.

If the fuel temperature increases, the mean free path increases (self-shielding decreases) due to the decreased microscopic cross section and more of the fuel pellet now sees a neutron flux with the energy level of the resonance peak. If the diameter of the fuel pellet is sufficiently large compared to the mean free path, the self- shielding effect is quite pronounced. - Even though the diameter to the fuel pellet may be 1 cm, not all paths lead through the center of the fuel pellet. The average straight line distance through a fuel pellet is about 0.625 cm. Reversing Example4-3, the three mean free paths yield a mean free path of 0.62513 or 0.21 cm. A 0.21 cm mean free path results in a microscopic cross section of about 240 barns. Therefore, in a real fuel pellet, any neutron at an energy level with a microscopic cross section of greater than 240 barns will, at some point, appear as a resonant energy neutron and be absorbed in the fuel pellet. - Refer to Figure 4-13 and examine the energy levels with cross sections above 240 barns. If the temperature increases to 600°F as in our previous example, the energy levels greatly expand with cross sections above 240 barns.

Therefore, the Doppler effect, when combined with the reduction in the self-shielding effect, results in increased resonance absorption at higher fuel temperatures. These examples support the U-238 cross sections pictured in Figure 4-13, but the affects of all resonant absorbers are similar.

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/ REACTIVITY COEFFICIENTS REV 3 i t 10.000 t 68°F 500 1,000 so& ) so; I 500 1,000 HIGH POWER 3.m TEMP (W 1.5N LOW POWER j". 3 +MODERATOR - COOLANT UEL ?k-ZlRCALOY - 4 CLADDING I $3 NEUTRON ENERGY (eV) Figure 4-13 U-238 Cross Section Curve The second issue that contributes to determining the amount self-shielding relates to the design characteristics of the fuel pellets. It studies how these design characteristics affect the actual temperature of a fuel pellet and how these factors combine to affect self-shielding.

t As previously stated, the fuel pellets are manufactured in the form of ceramic pellets.

Like any ceramic structure, they are poor conductors of heat. This causes a large temperature gradient from the center of the pellets to the outer surfaces. This is a major contributor to the reduction in self-shielding as fuel temperature increases.

Figure 4- I4 shows the temperature gradients encountered for fuel pellets in low and high power areas of the core.

In conclusion, by comparing the two gradient curves for the high and low temperature conditions for each 1°F increase in the average he1 temperature, the temperature gradients get larger between fuel centerline and the outer surfaces.

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/ REACTIVITY COEFFICIENTS REV 3 .) :.+ .-, uoz &HELIUM - PRESSURIZED GAP FUEL fl CENTERLINE Figure 4-1 4 Fuel Temperature Gradients For fuel pellets in the high powered regions of the core, fuel centerline temperatures above 3,000"F are common, while the temperatures near the fuel pellet surface may be =1,00O"F. For low power fuel pellets, the temperature gradients can range from 1,500"F centerline to 700°F at the surface.

Figure 4-15 illustrates the effect of the increasing temperature gradient on self- shielding.

Effective Area Shielded Effective Resonance Capture Region at , ,; -/From Resonance Capture

\. 'L -4 >Low Power uoz FGI Pellet at Low Power I Effective Area Shielded From Resonance Capture Effecbve Resonance Capture Region at /, _A >High Power UO? F~I Pellet at High Power Figure 4-15 Fuel Temperature Effects on SeV-Sh ielding An epithermal neutron that is at an off-resonance energy upon entering the low power pellet appears as a resonance energy neutron upon penetrating deeper into the pellet. Because the gradient is not as large as the high powered pellet. it can pass completely through the pellet and never be captured.

The same neutron entering the high powered pellet has a higher probability of appearing as a resonance energy neutron upon entering the pellet and a much greater probability of appearing as a resonance energy neutron as it goes deeper into the pellet.

Therefore, as fuel temperature increases, the effective capture area for epithermal neutrons also increases.

For the high powered pellets, only a very small fraction of the epithermal neutrons escape resonance capture because of the large increase in the effective capture area.

Why does increasing fuel temperature result in a greater fraction of neutrons in the core being resonantly captured even though Doppler broadening of the resonance peaks does not increase the probability that more neutrons will be lost? Because it is the combination of the Doppler broadening and fuel design. FUEL TEMPERATURE COEFFICIENT OR DOPPLER COEFFICIENT The Doppler coefficient (aD), also known as the fuel temperature coefficient. is defined as the change in reactivity per unit change in fuel temperature.

Where: aD = Doppler coeflicient (Ak/k/OF)

Ap = change in reactivity (Ak/k) = change in fuel temperature (OF) Equation 4-12 The Doppler broadening makes a larger fraction of the neutrons available for capture, even though the probability for capture does not increase.

The fuel design is such that it clumps a large volume of those probabilities (resonance absorbers) together in a very dense area, making it difficult for any one neutron to escape the probability of capture.

As fuel temperature increases, Doppler broadening adds a larger fraction of neutrons available for capture, and even if the probability for capture remains the same, more neutrons are absorbed because more are available to be absorbed.

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/ REACTIVITY COEFFICIENTS REV 3 A reactor with kern = 1.005 has a fuel temperature of 100°F. When fuel temperature is raised to 600°F.

bm = 1 .OOO. Calculate the value of the Doppler coefficient.

__ ~ Example 4-4 A good approximation for the Doppler coefficient is -1 x lo-' -OF. An increase in the fuel temperature results in a higher vibrational fiequency of the fuel atoms (increases Doppler broadening). The degree of self-shielding of the fuel is reduced, increasing the effective capture area of the fuel and resulting in a larger fraction of the neutrons in the core being resonantly captured.

In a low enrichment reactor, such as commercial reactors, most of the uranium in the fuel pins is U-238. The magnitude of the Doppler coefficient in these reactors is about -1 x 10-5 AWWOF. Although the coefficient is small in comparison with a,,, and a,, the reactivity effect increases to a very high value as the reactor goes from 0 to 100% power operation.

The peak fuel temperature in some fuel pellets could be as high as 4.000°F at 100% power. The average fuel temperature is about 1,200°F.

Thus, the reactivity effect due to the fuel temperature change is large. because the temperature change is large. ~ A reactor has an average Doppler coefficient of -0.8 x Ak/k/"F over the fuel temperature range from 100 to 1,600'F. Calculate the reactivity change associated with a fuel temperature change of 100 to 1,60O0F.

Example 4-5 The characteristic that makes the Doppler coefficient particularly important is that the fuel temperature immediately increases following an increase in reactor power. Since UO2 is a relatively poor conductor of heat and a cylindrical fuel pellet has a small heat transfer surface per unit volume, the time required is relatively long for the heat generated at any instant to be transferred to the moderator.

This required time is generally 7 to 9 seconds (shorter for newer fuels). BWR / REACTOR THEORY /CHAPTER 4 19 of 39 Q 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 In the event of a sudden and large positive reactivity addition to the reactor, the moderator temperature and void coefficients cannot interact for several seconds. Until the heat has a chance to be transferred from the fuel pellet to the moderator. the coefficients would have little immediate effect in countering the reactivity insertions.

The Doppler coefficient starts to react immediately and represents the primary shutdown mechanism for a fast power rise transient.

For this reason it is sometimes called the prompt coefficient, whereas the moderator and void coefficients are the delayed coefficients. The Doppler coeficient is one of the more important inherent safety features in low enrichment heterogeneous reactors.

CHANGES IN DOPPLER COEFFICIENT WITH CHANGES IN FUEL TEMPERATURE We have previously discussed, in depth, the Doppler effect and the effect that increasing fuel temperature has on the Doppler effect and resonance capture.

Therefore, the following discussion only focuses on the effect that increasing fuel temperature has on the magnitude (value) of the Doppler coefficient.

Each one degree increase in fuel temperature results in a smaller broadening of the resonant peaks, because as fuel temperature is increased, the atom movement is progressively more restricted due to the crystalline structure of the ceramic fuel pellets. This results in a progressively smaller fraction of epithermal neutrons available for resonance capture with each incremental increase in fuel temperature.

Consequently, as Figure 4-16 illustrates, as temperature increases, the magnitude of the Doppler coefficient decreases.

Figure 4-1 6 Doppler CoeBcient of Reactivity In Figure 4-16, a one degree Fahrenheit change from 500°F to 501°F results in a more negative value for the Doppler temperature coefficient than a 1°F change from 3,500"F to 2,501"F. This is because the additional vibration of the U-238 target nuclei is greater from 500°F to 501°F than from 2,500"F to 2,501"F. This results in a greater amount of Doppler broadening from 500°F to 501°F than from .~ 2,500"F to 2,501'F. It is important to note that aD is always negative.

Its negative magnitude is simply smaller in value at higher fuel temperatures.

/ CHANGES IN DOPPLER COEFFICIENT WITH CHANGES IN CORE AGE At the beginning of the fuel cycle, the fuel consists of U-238 and U-235. These fuels cause a reasonable amount of resonance absorption to occur. If the he1 temperature increases slightly, the broadening of the resonance peaks, primarily the U-238 peaks, causes a significant increase in the fraction of neutrons that are resonantly absorbed.

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/ REACTIVITY COEFFICIENTS REV 3 This results in a smaller fraction of thermal neutrons available for fission capture in the fuel. Therefore. the Doppler coefficient is negative at e BOC. At EOC. approximately the same amount of U-238 still exists. U-235 is reduced to about 60% of its original concentration, and Pu-239 and Pu-240 have significantly increased in the core. i As fuel is used over core life, the following reactions produce Pu-240.: Equation 4-13 239~ p',r 239 92 + 93NP fIl2 = 23.5m Equation 4-14 Equation 4-15 Equation 4-1 6 Note that Pu-239 produces Pu-240 by way of neutron capture about 27% of the time.

Neutron absorption in Pu-239 results in fission about 73% of the time.

Figure 4-17 presents a total macroscopic cross section graph of Pu-240.

Note that the total cross section shown is mostly for that capture.

SLOW INTERMEDIATE FAST (EPITHERMAL) 1,mm c I I 100 OOO lo.m I 1.m 100 e 10 P, 1 10 Figrrrv 4- I' Pu-,$40 Total Neutron

('rm s Section As a result ot I'u-240 production over core life. the Dopplcr tcriipc.r.rlurc coefficient becomes more ncpti\c t-wc.iiiw f'u-240 has a very high capturc crohs w*cliori tiir I eV kinetic energy incident ncutnw. nmicly about 1 x IO'bams. Thereforc.

ah l'u-240 huilds up, the value for aD becomes mcrrc ncpti\c later in core life as shown in f..igurc 1- 18 Fission product3 arc present that were not present at IN I('. These materials resonantly capture a sizcahlc number of neutrons. The major contributors to the Doppler coefficient are U-238 and Pu-240. A small fuel temperature increase at EO(' causes the broadening of the U-238 peaks a prc\hdy described.

with thc addition of thc l'u-240 peaks to broaden. The presence of Pu-240. along with the extra fission products with high resonance absorption peaks. causes a large fractional increase in the number of neutrons undergoing resonance capture.

Figure 4-1 8 shons that the Doppler coefficient becomes morc negative due to the buildup of additional resonant absorbers as the core ages. u BWR / REACTOR THEORY /CHAPTER 4 21 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 CHANGES IN DOPPLER COEFFICIENT WITH CHANGES IN MODERATOR DENSITY - If the moderator density is high (low temperature), the slowing down length and the slowing down time of a neutron are relatively short. When the moderator is hot or contains voids, the slowing down length and the slowing down time for neutrons are longer and any change in the resonance peaks are more significant since the neutrons can travel farther and spend relatively longer periods of time in the resonance region.

As Figure 4-1 8 shows, the Doppler coefficient is more negative at high moderator temperatures and is most negative at high void fractions. AVERAGE FUEL TEMPERATURE (OF) 500 1m 15M) zoo0 2500 Jwo 3500 4am 4500 iz 021 II Ill I1 I I a 7 Figure 4-1 8 Doppler Coefficient of Reactivity Figure 4-1 8 helps depict the Doppler coefficient. The Doppler temperature coefficient (aD) is always negative throughout core life. The Doppler temperature coeficient (010) becomes more negative with core age and as moderator density decreases.

As fuel temperature increases, CLD becomes less negative.

An average value for aD is: Aklk aD = -1 x IO-'- "F Equation 4-1 7 The Doppler coefficient is the first coefficient to respond to an accidental. large, positive reactivity addition. The Doppler coefficient's importance becomes paramount in the event of a cold water accident or an ejected rod accident.

If core power increases rapidly, fuel temperature increases and a large time lag exists before the transfer of heat to the moderator (seven seconds or more). As fuel temperature increases, more and more negative reactivity is added to the core to counteract the reactivity addition. ~~ ~~~ ~~~~ ~ During a reactor coolant system cooldown, positive reactivity is added to the core (assuming a negative moderator temperature coefficient).

This is mainly due to: a. an increase in the resonance escape probability.

b. a decrease in the resonance escape probability . c. an increase in the thermal utilization factor. d. a decrease in the thermal utilization factor. Example 4-6 : c2# BWR / REACTOR THEORY / CHAPTER 4 22 of 39 0 2000 GENERAL PHYSICS CORPORATlON

/ REACTIVITY COEFFICIENTS REV 3 c c more neg then less nex no change aV Which of the following best describes how Doppler broadening of resonance absorption peaks contributes to making the fuel temperature (Doppler) coefficient of reactivity negative?

As fuel temperature increases:

more ncg a. b. C. d. the absorption cross section for the resonance peaks increases.

causing more absorption of resonant energy neutrons.

absorption of off-resonance neutrons increases while absorption of resonant energy neutrons remains relatively constant. resonance energy absorption cross sections decrease, resulting in increased resonance escape. the neutron energy spectrum is "hardened", resulting in more resonance absorption.

Example 4-7 Complete the following matrix regarding more negative, no change, or less negative:

I Vote I: KO change during normal power operations.

Example 4-8 Calculate the stable reactor period thal results from the collapse of 1% of the voids in a reactor at 100% power. assuming:

pcff = 0.006 - 3C = A,, = 0.1 sec-' a, = -1 x Aklk %voids -~ Example 4-9 BWR / REACTOR THEORY

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/ REACTIVITY COEFFICIENTS REV 3 L The void fraction in a BWX increases from 25% to 35%. If the void coefficient is -3.5 x MU% voids, reactivity inserted.

calculate the ~ Example 4-1 0 -~ List the approximate values for the Doppler, moderator, and voids reactivity coefficients.

&ample 4-11 I POWERCOEFFICIENT I I I It is comeiiicnt to combine the various reacthit!

cocl'licicnts into a single coefficient.

Although the cc~cl'hcicnts are associated with fuel tenipcrarurc. niodcrator temperature, and voids. ultinutcl!

thc quantity of concern is reactor pcwcr. Kcactor power is easily measurclhlc 123 oyp)scJ to YO voids or fuel temperaturc I ;ird thc rcactivity changes due to changes in reactor power can be readily calculated.

The definirion

(,I' pwer coeficient is in a manner a1~111ytu~

to other reactivity coefficicnls: - - AP AO/O Power Eqrration 4 For practical purpscs. the only coefficients - considered arc 1hc void coefficient and the fuel temperature coefficient.

Once the moderator is at normal operating temperature, it does not change significantly from 0% power to 100% power in a BWR. The power coefficient can be rewritten as:

a ,, AT,uc, + a A%voids - a Pmbcr - AYo Power Equation 4-1 9 When analyzing reactor transient response.

it is important to know how the reactivity coefficients respond to a transient. The three transient classes are:

pressure, water inventoryhernperature changes, and power. The following discussion identifies the first coefficient that responds to the transient.

BWR / REACTOR THEORY

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/ REACTIVITY COEFFICIENTS REV 3 c L For pressure transients, the first reactivity coefficient to respond will be the void coefficient of reactivity.

When a pressure increase occurs (as in an inadvertent main steam line isolation or turbine tip). the voids in the core collapse. This appears as a large decrease in YO voids and as a large positive reactivity insertion due to the increase in thermal neutrons resulting fiom the density increase of the moderator.

Reactor power increases.

For a depressurization transient (as in a steam line break or safety relief valve lifting), the voids in the core expand due to the pressure drop. This appears as a large increase in %voids and a large negative reactivity insertion; subsequently, neutron moderation decreases, resonant absorption increases, and reactor power decreases.

Water inventory/temperature decrease transients can be caused by either an injection of cold water from an inadvertent Emergency Core Cooling System (ECCS) initiation, a loss of Feedwater (FW) heaters due to a turbine trip, or loss of extraction steam.

When colder water enters the core, moderator temperature drops, some voids collapse, and the density of the moderator increases resulting in an increase of neutron moderation and a decrease in resonant absorption. This appears as a large positive reactivity increase, resulting in a reactor power increase.

Either a loss of shutdown cooling, loss of feedwater, or loss of level causes water inventory/temperature increase transients.

When an increasing transient in water inventory/temperature occurs, warmer water enters the core. This causes moderator temperature to increase and the density of the moderator decreases. This results in a decrease in neutron moderation.

an increase in resonance absomtion.

and a decrease in reactor Dower. For power transients (as in a rod drop or inadvertent rod withdrawal), power increases due to increased fuel added to the effective core (or decreased absorption in control rods) and leads to an increase in the number of fissions.

which increases the amount of energy released in the fuel. The fuel temperature will increase rapidly. The fuel thermal time constant limits the removal of heat from the fuel by the moderator.

As fuel temperature rapidly increases, the Doppler coefficient adds negative reactivity.

More neutrons are lost to resonance absorption and reactor power begins to turn (rate of increase slows and then decreases).

In reactor design.

it is essential that both the void coefficient and he1 temperature coefficient are negative.

If power increases due to a positive reactivity insertion, the resultant increase in fuel temperature and void fraction adds negative reactivity, which in turn limits or turns the power increase. This phenomenon makes the reactor inherently stable due to a negative reactivity feedback effect. If these coefficients are positive, an increase in reactivity produces an increase in power that in turn adds positive reactivity, and the reactor can "run away". Chernobyl Unit 4 is an example. Chernobyl was designed to have a positive moderator/void coefficient. Therefore, as the water in the reactor coolant began to heat up and create voids during that incident in 1986, a large positive reactivity was inserted.

This rendered the reactor prompt supercritical, which destroyed the reactor.

Due to the large magnitude of the void coefficient, the power coefficient is stronger at higher power levels. Typical values for the power coefficient are in the range of -0.03% AWYO power to -0.06% WYO power. I ',d BWR / REACTOR THEORY / CHAPTER 4 25 of 39 0 2000 GENERAL PHYSICS CORPORATION I' REACTIVITY COEFFICIENTS REV 3 I REACTIVITY DEFECTS I The term "reactivity defect" (px) is used to describe the total amount of reactivity added, positive or negative, due to changing a parameter by a given amount. For example:

P, =(AxXa,) Where: PX X Ax QX = reactivity defect (Akk) = specific parameter

(% voids, fuel temp, moderator temp)

= change in parameter x

= parameter x reactivity coefficient (YO voids, fuel temp, moderator temp)

Equation 4-20 A reactor operating with a void coefficient of -1 x Ak/k/?h voids undergoes a pressure increase that causes a 4% decrease in void fraction. Calculate the reactivity added. Example 4-12 Another example of a reactivity defect involves the Doppler defect.

For example, changes in fuel temperatures affect reactivity of the core and subsequently kn. The effects of fuel temperature change on the factors in the six factor formula will now be discussed.

A fuel temperature increase does not affect the reproduction factor (11). fast fission factor (E), both nonleakage terms (Jf and -&). and thermal utilization factor (f). Therefore, as more neutrons are resonantly absorbed, the resonance escape probability (p) decreases.

Also, as fuel temperature increases, resonance absorption increases, and p decreases.

For the BWR core, as fuel temperature and power increase, negative reactivity will be always inserted. This results in a negative effect on power and br. In fact, because the effect of resonances is occurring at the source of fission (the fuel), Doppler will be the quickest negative reactivity insertion to help turn a power upswing or power excursion. This negative reactivity insertion as a function of power is referred to as the Doppler defect. Figure4-19 shows an example of how the Doppler defect behaves.

RATED POWER (X) Figure 4-19 Doppler Defect BWR i REACTOR THEORY / CHAPTER 4 26 of 39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 Calculate the total reactivity addition to the core for a transient where fuel temperature rises 400°F, moderator temperature rises 2OF, and void fraction rises 2%. given the following:

Akfk a,,, = -1.5 x 10-'0 Frnd Akfk %voids a, =-I.IXIO-~

Example 4-13 A BWR is operating at full power with a 38% void fraction and an effective fuel temperature of 1.350"F. Determine the reactivity defects for voids and Doppler assuming a, = -1 x M% voids and ag = -1 x AWW'F. Also determine what fraction of the total defect is due to voids. Assume that True, at 0% power = 550°F and rods remain at their designated 100% rod pattern. Example 4-14 ~~ BWR / REACTOR THEORY /CHAPTER 4 27 of 39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 REACTIVITY BALANCE AND DESIGN I CONSIDERATIONS A convenient method of summarizing a reactor's state of criticality or overall reactivity value is by performing a reactivity balance. In reality, these balances are numerically inaccurate, but they provide a gross indicator to the operator.

All such balances assume for a starting point that the reactor is capable of achieving cold criticality. Therefore a cold, clean reactor with a k,=n = 1.0 is the starting point.

Cold, as used here, is 68°F and clean means that there are no fission produced poisons in the reactor, i.e., xenon, samarium, etc. After initial criticality and power operation, the reactor is no longer considered clean. The reactor is then referred to as "xenon-free" rather than "clean." Since our starting point is from a cold, clean, critical condition, enough fuel (positive reactivity) must be assembled to form a critical mass at 68°F. Much more fuel (positive reactivity) must be added to the critical mass to achieve 100% power equilibrium conditions.

The term for this extra added reactivity above the amount for a critical mass under cold, clean conditions is excess reactivity (pex). The effective multiplication factor (ken) associated with this excess reactivity is termed kexcess. The definition of ~xccss is the amount of neutron multiplication available above that required for criticality.

The maximum effective multiplication factor (kmax) will equal hff under the following conditions: cold, clean, no control rods inserted The definition of kmax is the maximum amount of neutron multiplication available under these conditions. The value of kern listed as kmax is. in effect. the value for the installed value of kern at beginning of life (BOL) conditions.

If k,, is known. the excess reactivity can be calculated by using the formula for determining reactivity when the effective neutron multiplication factor is known. Therefore:

T Equation 4-22 For this case, we define the effective neutron multiplication factor as k,, and the reactivity as excess reactivity. This results in the following relationship: Equation 4-23 To determine how much more fuel (positive reactivity) must be added to achieve 100% power at equilibrium conditions, we must examine what transpires in going from the cold, clean, critical condition to a 100% power equilibrium condition.

k,,,, = k, - 1 = k,, - 1 Equation 4-21 BWR / REACTOR THEORY

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/ REACTIVITY COEFFICIENTS REV 3 First. we have to heat up from cold to a hot operating temperature of about 550°F. When this is done. we need to know how much negative reactivity is added because of the moderator and Doppler coefficients. The reactivity defect associated with the moderator temperature increase is: ( p, =a,AT AT = 550 - 68°F = 482°F p, = -4.82%Ak / k Equation 4-24 The reactivity defect associated with the fuel temperature increase to an average operating fuel temperature of 1,400"F is: pD =a,AT c AT = 1400 - 68°F = 1332°F pD =-1.332%Aklk Equation 4-25 The reactivity defect associated with the increase in voids from 0% to 38% is: pv = a,A%voids A %voids = 38% - 0% = 38% -l~lO-~ Ak/k )(38%voids)

%voids pv =-3.8%Ak/

k Equation 4-26 Knowing that if there is only enough fuel in the reactor to achieve a cold. clean.

critical mass. raising the temperature would make the reactor subcritical (k~ < 1 .O). Therefore, positive reactivity must be added to stay critical.

The addition of positive reactivity in the form of excess fuel keeps the reactor critical.

This amount of positive reactivity, according to our reactivity balance equation. must be equal to the negative reactivity added by heating up to 550°F. or +4.82 %Ak'k plus +I .332 %Am plus +3.8% %AWk. which totals +9.952% Auk. This value represents the reactivity that must be present to maintain the reactor critical assuming it is still in a "clean" condition. However, the production of fission product poisons begins as soon as the reactor is critical.

The next step is to reach equilibrium conditions at 100% power. In this case, we refer to equilibrium xenon (Xe) and samarium (Sm). Equilibrium Sm = -1.0% Ak/k Equilibrium Xe = -3.0% Ak/lc As these poisons build up in the reactor, sustained criticality would be impossible if we did not add enough fie1 to compensate. Therefore, to remain critical, we must again balance this reactivity by adding enough fuel to equal +4% Akk. We have now added enough fuel to operate at the 100% power equilibrium conditions.

As we operate, the he1 in the reactor depletes. Fuel depletion adds negative reactivity and the reactor will become subcritical.

As this happens, power decreases, causing the fuel temperature and moderator temperature to decrease, adding positive reactivity to the reactor. This positive reactivity addition offsets the negative reactivity from fuel depletion and the reactor will become critical again but at a lower power level.

Since we need to generate power, we do not want this to occur. To allow operation for a specified amount of time, we add even more fuel than required just to get to the ',I BWR / REACTOR THEORY /CHAPTER 4 29 of 39 0 2000 GENERAL PHYSICS CORPORATION

' REACTIVITY COEFFICIENTS REV 3 c 100% power equilibrium conditions.

This specified time is the he1 cycle. An 18-month fuel cycle requires approximately

+ 15% AWk. ~ Calculate the excess reactivity (fuel) added to the reactor to operate at 100% power for 18 months. Example 4-15 . -4 d BWR / REACTOR THEORY

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,' REACTIVITY COEFFICIENTS REV 3 Control Rod Density (CRm Doppler Broadening Doppler Coefficient or Fuel Temperature Coefficient (aD) Fuel Temperature GLOSSARY The percentage of the control rods inserted into the core where 100% control rod density implies all rods are fully inserted. The widening and flattening effect on resonance capture probability peaks for epithermal neutrons due to increased kinetic energy of target atoms resulting from increased fuel temperature. The reactivity coefficient that relates the change in reactivity due to a change in fuel temperature. It is given as aD = Ap/ATf,,l and has units of W"F. When discussing the Doppler temperature coefficient or fuel temperature coefficient, some facilities use "average fuel temperature" while others use "effective fuel temperature."

CENTERLINE FUEL TEMPERATURE AVERAGE FUEL TEMPERATURE EFFECTIVE FUEL TEMPERATURE EDGE OF PELLET FUEL 1 /TEMPERATURE EDGE OF PELLET I AXIAL CENTERLINE The "average fuel temperature

is an average between the centerline and the edge of the pellet. The "effective fuel temperature" denotes the he1 temperature where most of the resonance capture "effectively" occurs. Typically, this can be somewhat lower than the average. BWR i REACTOR THEORY / CHAPTER 4 31 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 L Moderator Temperature Coefficient (a,) Power Coefficient (ap) Reactivity Defect (px) Resonance Energy Self-shielding GLOSSARY The reactivity coefficient that relates the change in reactivity due to a change in moderator temperature.

It is given as a,,, = Ap/ ATmod and has units of AkW'F. The reactivity coefficient that relates the change in power to the total effect of moderator temperature coefficient.

void coefficient.

and Doppler coefficient.

It has units of Ak/k% power and is given as: - a, AT, + a,A%voids AYo Power a,o*,, - The moderator coefficient is often omitted because of the small changes that occur in moderator temperature once the BWR reactor is at power operation. The total amount of reactivity, positive or negative, due to a changed plant parameter.

It is given as:

(3 p, = (Ax)(a,)= (Ax There exist discrete excitation energy levels within a nucleus (such as U-238 or Pu-240). If the incident neutron kinetic energy is equal to one of these excitation energy states, the neutron is said to be at a resonance energy for that nuclide. Note that as the target nucleus vibrational energy increases, the range of neutron energies broadens where the relative energy between the incident neutron and target nucleus is equal to one of these resonance energies.

The phenomenon where resonant energy level neutrons are absorbed in the outer layers of a fuel pellet, thereby never being absorbed in the central areas of the fuel. Therefore, the outer layers shield the inner layers and the pellet is said to be self- shielded.

4 BWR / REACTOR THEORY / CHAPTER 4 32 of 39 0 2000 GENERAL PHYSICS CORPORATION i) REACTIVITY COEFFICIENTS REV 3 EXAMPLE ANSWERS _r L c A reactor operating at 530°F has a ktT= 1.000. The moderator temperature is increased to 540°F and k,r decreases to 0.999. Calculate the value of the moderator temperature coefficient.

Solution:

= -1.001 x IO-) - (0.999 - 1) 0.999 Pfinal - (- 1.001 x 10-~)-(0) a, = 540 - 530°F - 1.001 x 10-~~k/k a, = 1 OOF AkIk a, = -1.001 x 10-~ - OF Example 4 A reactor has ;in mwrlge core void fraction of 2O0,o uith h,,, - 1.000. Calculate the void coc tlicic*iit I I' the void fraction is increased IO 2_"'t8 rcwlting in a kfi- decrease to 0.0ox (- 2 x lo-q)-(o) a, = (X""\oids-20%voids)

-2 x IO 'Ak/k 2 " 01 oids a, = . Akik a, =-I x 10.' O/bvoids &ample 4-2 BWR / REACTOR THEORY /CHAPTER 4 33 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 Calculate the mean free path of the 21 eV neutron and the value of three mean free paths. 1 ha =- No, 1 = (2 x 10" -)(5500 atoms barns)(lO~"'

K) cm' barns ha = 0.009cm Therefore, three mean free paths are equal to: (3)Aa = (3)(0.009 cm) = 0.027 cm ~ Example 4-3 A reactor with LR = 1.005 has a fuel temperature of 100°F. When fuel temperature is raised to 600°F. Lm = 1 .OOO. Calculate the value of the Doppler coeficient.

(0)- (4.98 x 10-3Ak/ k) a, = 600 - 100°F -4.9gX 10-~~k/k a, = a, = -9.96~ 1O4Ak/k/OF 500°F ~~ ~ Example 4-4 \ -z-d r BWR / REACTOR THEORY

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/ REACTIVITY COEFFICIENTS REV 3 EXAMPLE ANSWERS L ~~ ~ ~~~ ~ ~~~ A reactor has an average Doppler coefficient of -0.8 x W"F over the fuel temperature range from 100 to 1,600"F. Calculate the reactivity change associated with a fuel temperature change of 100 to 1.600"F. Solution:

AP AT,,, a, =- Ap = (- 0.8 x (1600- 10O0F) During a reactor coolant system cooldown, positive reactivity is added to the core (assuming a negative moderator temperature coefficient). This is mainly due to: a. an increase in the resonance escape pro babi I i t y . b. a decrease in the resonance escape probability . c. an increase in the thermal utilization factor. d. a decrease in the thermal utilization factor. Answer: a Example 4-6 '4 BWR / REACTOR THEORY

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/ &EACTIVITY COEFFICIENTS REV 3 L 7 Which of the following best describes how Doppler broadening of resonance absorption peaks contributes to making the fuel temperature (Doppler) coefficient of reactivity negative?

As fuel temperature increases:

a. b. C. d. the absorption cross section for the resonance peaks increases, causing more absorption of resonant energy neutrons.

absorption of off-resonance neutrons increases while absorption of resonant energy neutrons remains relatively constant.

resonance energy absorption cross sections decrease, resulting in increased resonance escape. the neutron energy spectrum is "hardened, resulting in more resonance absorption.

Answer: b Example 4-7 Complctc ttic Iidhw ing matrix regarding more nt'gati\.c.

no change, or less negative:

Ansiver: IT\\ more nc change' \ole 1: ho chvngr during normal power operations.

Example 4-8 BWR / REACTOR THEORY

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/ REACTIVITY COEFFICIENTS REV 3 1 EXAMPLE ANSWERS i I m f Calculate the stable reactor period that results from the collapse of 1% of the voids in a reactor at 100% power. assuming:

p,, = 0.006 Ak/k %voids a, = -1~10-~ Solution:

p=Ak/k=a, x%voids 3 &Ik )(- I%voids) %voids p = 1 x lo-' Ak/ k = 0.001 Ak/ k Therefore, the stable period would be: - pew - p 0.006 - 0.001 T= - hp =wj T = 50 seconds Example 4-9 ~~ The void fraction in a BWR increases from 25% to 35%. If the void coefficient is -3.5 x Ak/k/% voids, calculate the reactivity inserted.

Solution:

%change in voids = 35% - 25% = 10% Ap = a, x %change in voids %voids )(I O%voids) *' Ap = -3.5 x 1 0-' Ak / k ~~~ Example 4-10 BWR / REACTOR THEORY / CHAPTER 4 37 of 39 C 2000 GENERAL PHYSICS CORPORATION

,' REACTIVITY COEFFICIENTS REV 3

- EXAMPLE ANSWERS List the approximate values for the Doppler, moderator, and voids reactivity coe ff c ien ts. Answer: Ak/k a, s -1 x IO-' - "F Aklk a, = -1 x10- - OF Aklk %voids a, z-IxIO-~ Note: When written in this order, their powers of ten are -5, -4, and -3 and refer to D, m, and V, respectively.

They can be remembered by using the following mnemonic device: Department of - Motor - Vehicles.

Ekample 4-11 A reactor operating with a void coefficient of -1 x AW% voids undergoes a pressure increase that causes a 4% decrease in void fraction. Calculate the reactivity added. p, = (A%voidsXa,)

%voids p, = (- 4%voids pv = 0.004Ak/k Positive reactivity was added.

Example 4-I2 Calculate the total reactivity addition to the core for a transient where fuel temperature rises 400°F. moderator temperature rises 2°F. and void fraction rises 2%. given the following:

a, =-1.3xlO-'-

Aklk OFfuet Aklk a, =-lSxlO-'-

OFmOd Aklk %voids a, =-1.1x10-3

%voids pIotal = (- 0.0052Ak / k)+ (- 0.0003Ak / k) + (- 0.0022Ak I k) pIotal = -0.0077Ak

/ k ~ &ample 4-13 ./ BWR / REACTOR THEORY / CHAPTER 4 38 of39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTlVlTY COEFFICIENTS REV 3 I EXAMPLE ANSWERS f' f A BWR is operating at full power with a 38% void fraction and an effective fuel temperature of 1,350"F. Determine the reactivity defects for voids and Doppler assuming av = -I x Auk/% voids and an = -1 x 1 0-5 AW"F. Also determine what fraction of the total defect is due to voids. Assume that Tfuel at 0% power = 550°F and rods remain at their designated 100% rod pattern. Solution:

x 38%voids Ak/k YO voids = -1 IO-' Apv = -3.8 x 1 0-2 Ak / k And APD = ApD =-8~10-~Ak/k Total reactivity defect due to v Doppler is:

ids Ak - (- 3.8 x 1 O-?)+ (- 8x 1 0-3)- k APvm - d Example 4-14 (Continued in next column) f i. Fraction of total reactivity defect at 100% reactor power due to voids is then approximately:

-3.8~10-'Aklk bV&" -4.6~10-'Ak/k

= 0.826 APV - - Fraction of total reactivity defect at 100% reactor = 83% power due to voids Example 4-14 ~ Calculate the excess reactivity (fuel) added to the reactor to operate at 100% power for 18 months. 4.820% Akk due to a,,, (heat up from 68" to 550°F) 1.332% AMC due to aD (heat up from 68" to 1,400"F) 3.800% Akk due to av (change 0% to 38%voids) 1 .OOO% Akk due to Samarium 3.000% Akk due to Xenon 15.000% Akk for 18 month cycle 28.952% Akk excess reactivity To simplifj calculations, round off this number to +29% Akk excess reactivity.

Example 4-15 BWR / REACTOR THEORY / CHAPTER 4 39 of 39 0 2000 GENERAL PHYSICS CORPORATION

/ REACTIVITY COEFFICIENTS REV 3 L2 C11 MICROBURN 82 MTC information Temp 68 170 250 350 450 Exposure MwdMt Groups K-eff-68 K-eff-170 K-M-250 K-eff-350 K-eff-450 Delta (68-68) Delta (170-68) Delta (250-68) Delta (350-68)melta (450-68) -0.02308 -0.03404 0 ARI 0.94443 0.93858 0.93153 0.92135 0.91039 0.00000 -0.00585 -0.01290 0 GPR 1 Q48 0.98589 0.98157 0.97615 0.96734 0.95517 O.Ooo00 -0.00432 -0.00974 -0.01855 -0.03072 -0.01855 -0.02946 0 GPR2 Q48 1.00379 0,99944 0.99392 0.98524 0.97433 0.00000 -0.00435 -0.00987 -0.01 667 -0.02692 0 GPR 4 Q 48 1.02718 1.02333 I .01842 1.01051 1.00026 0.00000 -0.00385 -0.00876 0 GPR 3 Q 48 1.04213 1.03892 1.03475 1.02785 1.01845 O.Ooo00 -0.00321 -0.00738 -0.01428 -0.02368 0.000 -0.002 -0.004 -0.006 4.008 -0.010 -0.012 -0.014 % -0.016 c, -0.020 (0 -0.018 Q) -0.022 p -0.024 - -0.026 -0.028 -0.030 -0.032 -0.034 -0.036 -0.038 -0.040 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 380 400 420 440 460 480 Moderator Temperature (F) L2 C11 MB2 Delta K vs Moderator Temperature (at BOC) + ARI +GRPl@ 48 +GRP2@48 4 GRP4 @ 48 U GRP3 @ 48 L2 Cl 1 MICROBURN 82 MTC information Temp 68 170 250 350 450 Exposure MwdlMt Groups K-eff-68 K-eff-170 K-eff-250 K-eff-350 K-eff-450 Delta (68-68) Delta (170-68) Delta (250-68)

Delta (350-68)elta (450-68) 1.9 ARI 0.93681 0.93169 0.92549 0.91651 0.90660 0.00000 -0.00512 -0.01 132 -0.02030 -0.03021 1.9 GPR I @48 0.97826 0.97487 0.97043 0.96259 0.95091 0.00000 -0.00339 -0.00783 -0.01567 -0.02735 -0.01565 -0.02544 1.9 GPRZ @48 0.99487 0.99122 0.98663 0.97922 0.96943 0.00000 -0.00365 -0.00824 1.9 GPR4@48 1.01762 1.01451 1.01053 1.00392 0.99494 0.00000 -0.0031 1 -0.00709 -0.01370 -0.02268 1.9 GPR 3 @48 1.03243 1.02997 1.02672 1.02109 1.01295 0.00000 -0.00246 -0.00571 -0.01134 -0.01948

-0 028 - -0030 - LZCII MB2 Delta K vs Moderator Temperature - \-,. m w f ARI GRPl@ 48 -A-GRP2@48 4GRP4@48 _I (at 01 900 Mwd/Mt) I UGRP3@48 L2 Cl 1 MICROBURN 82 MTC information Temp 68 170 250 350 450 Exposure MwdMt Groups K-eff-68 K-eff-170 K-eff-250 K-eff-350 K-eff-450 Delta (68-68) Delta (170-68) Delta (250-68) Delta (35068)*elta (45068) 6.8 ARI 0.92750 0.92421 0.92016 0.91382 0.90557 0.00000 -0.00329 -0.00734

-0.01368 -0.02193 -0,01049 -0.01961 6.8 GPR 1 @48 0.96562 0.96352 0.96065 0.95513 0.94601 0.00000 -0.00210 -0.00497

-0.00856 -0.01639 -0.00594 -0.01301 6.8 GPR4e.48 1.00506 l.OO408 1.00257 0.99912 0.99205 0.00000 -0.OOO98 -0.00249 6.8 GPR3@48 1.01920 1.01880 1.01801 1.01555 1.00936 0.00000 -0.o0040 -0.001 19 -0.00365 -0.00984 6.8 GPR 2 @ 48 0.98277 0.98104 0.97873 0.97421 0.96638 0.00000 -0.001 73 -0.00404 -B- GRPl Q 48 L2CI I MB2 Delta K vs Moderator Temperature +GRP2@48 +GRP4@48 (at 06800 MwdlMt) , 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 380 400 420 440 460 480 Moderator Temperature (F) (Intersection of Lines on X Axis at 68 Degrees F)

L2 C1 IMICROBURN 82 MTC information Temp 68 170 250 350 450 Exposure Mwd/Mt Groups K-eff-68 K-eff-170 K-&-250 K-&-350 K-eff-450 Delta (68-68) Delta (170-68) Delta (25U68) Delta (350-68))elta (450-68) 8.55 ARI 0.92822 0.92571 0.92248 0.91702 0.90894 0.00000 -0.00251 -0.00574 -0.01 120 -0.01928 -0.00347 -0.00796 -0.01656 8.55 GPRl @48 0.96501 0.96361 0.96154 0.95705 0.94845 0.00000 -0.00140 8.55 GPRl Q 48 0.98268 0.98179 0.98039 0.97690 0.96931 0.00000 -0.00089 -0.00229 -0.00578 -0.01 337 -0.00305 -0.00987 8.55 GPR4Q48 1.00504 1.00488 1.00433 1.00199 0.99517 0.00000 -0.00016 8.55 GPR3Q48 1.01894 1,01940 1.01960 1.01823 1.01230 0.00000 O.OOO46 O.OOO66 -0.00071 -0.00664 -0.00071 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 380 400 420 440 460 480 Moderator Temperature (F) (Intersection of Lines on X Axis at 68 Degrees F)

L2 C1 1MICROBURN B2 MTC information Temp 68 170 250 350 450 Exposure Mwd/Mt Groups K-eff-68 K-eff-170 K-eff-250 K-eff-350 K-eff-450 Delta (68-68) Delta (1 70-68) Delta (250-68) Delta (350-68)slta (450-68) 10.8 ARI 0.93186 0.93027 0.92801 0.92343 0.91512 0.00000 -0.001 59 -0.00385 -0.00843 -0.01674 10.8 GPRl @48 0.96736 0.96708 0.96611 0.96267 0.95397 0.00000 -0.00028 -0.00125 -0.00469 -0.01339 10.8 GPR2 a48 0.98539 0.98553 0.98527 0.98279 0.97509 0.00000 0.00014 -0.00012 -0.00260 -0.01030 10.8 GPR4@48 1.00769 1.00865 1.00932 1.oO808 1.00110 0.00000 O.OOO96 0.00163 0.00039 -0.00659 10.8 GPR3@48 1.02142 1.02300 1.02439 1.02415 1.01808 0.M)oOo 0.00158 0.00297 0.00273 -0.00334 +GRP2@48 +GRP4@48 U GRP3 @ 48 LZCII MB2 Delta K vs Moderator Temperature (at 10800 Mwd/Mt) 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 360 380 400 420 440 460 480 Moderator Temperature (F) (Intersection of Lines on X Axis at 68 Degrees F) - _- -

L2 Cl 1 MICROBURN 82 MTC information Temp 68 1 70 250 350 450 Exposure MwdMt Groups K-eff-68 K-eff-170 K-eff-250 K-eff-350 K-eff-450 Delta (68-68) Delta (170-68) Delta (250-68) Delta (350-68)slta (450-68) 13.5 ARI 0.93653 0.93563 0.93403 0.92992 0.92112 0.00000 -0.00090 -0.00250 -0.00661 -0.01541 13.5 GPR 1 Q 48 0.97087 0.97130 0.97111 0.96826 0.95914 0.00000 0.00043 0.00024 -0.00261 -0.01 173 13.5 GPR2Q48 0.98884 0.98985 0.99035 0.98853 0.98040 0.00000 0.001 01 0.00151 -0.00031 -0.00844 13.5 GPR4Q48 1.01095 1.01273 1.01422 1.01365 1.00625 0.00000 0.00178 0.00327 0.00270 -0.00470 0.00468 0.00503 -0.00132 13.5 GPR3 Q 48 1.02432 1.02673 1.02900 1.02935 1,02300 0.00000 0.00241 LZCl1 MBZ Delta K vs Moderator Temperature (at 13500 Mwd/Mt) f ARI + GRPl @! 48 GRP4 @ 48 +GRP2 @! 48 GRP3 d 48 Moderator Temperature (F) (Intersection of Lines on X Axis at 68 Degrees F)

L2 C1 IMICROBURN B2 MTC information Temp 68 170 250 350 450 Exposure MwdlMt Groups K-eff-68 K-eff-170 K-eff-250 K-eff-350 K-eff-450 Delta (68-68) Delta (170-68) Delta (250-68) Delta (350-68)selta (450-68) 17 ARI 0.93318 0.93231 0.93068 0.92635 0.91670 0.00000 -0.00087 -0.00250 -0.00683 -0.01648 17 GPRl @48 0.96544 0.96597 0.96583 0.96287 0.95297 0.00000 0.00053 0.00039 -0.00257 -0.01247 17 GPR2 Q48 0.98436 0.98536 0.98587 0.98378 0.97469 0.00000 0.00100 0.00151 -0.00058 -0.00967 17 GPR4@48 1.00621 1.00805 1.00951 1.00867 1.OOO30 0.00000 0.001 84 0.00330 0.00246 -0.00591 0.00524 -0.00214 17 GPR 3 Q 48 1.01804 1.02064 1.02295 1.02328 1.01590 0.00000 0.00260 0.00491 L2Cl1 MB2 Delta K vs Moderator Temperature (at 17000 Mwd/Mt) 0.008 0.006 0.004 0.002 0.000 y -0.002 (D -0.004 5 -0.006 4.010 -0.012 4.014 -0.016 -0.018 Y P -0.008 t ARI -t GRPI @ 48 -A-GRP2@48

+GRP4@48 -U-GRP3a48 Moderator Temperature (F) (Intersection of Lines on X Axis at 68 Degrees F)

BWR GENERIC FUNDAMENTALS REACTOR THEORY CHAPTER 2 NEUTRON LIFE CYCLE i 1 27 THERMAL NEUTRON LEAKAGE 326 THERMAL NEUTRONS ABSORBED BY 427 RESONANCE NON-FUEL ATOMS 1800 FAST MODERATOR U-235 FUEL

+ill+ START CYCLE 74 HERE FAST NEUTRON LEAKAGE NEUTRONS FROM FAST f ISSION STUDENTTEXT REV 3 62000 General Physics Corporation, Columbia, Maryland All nghts reserved No put of &is book may be rcpmduccd in any form 01 by my me-. wihuc permission in wnhng from General Physss Caporlhon

. TABLE OF CONTENTS .. TABLES AND FIGURES ............................................................................................................

11 OBJECTIVES

..............................................................................................................................

111 K/A - OBJECTIVE CROSS REFERENCE

................................................................................

iv STEADY STATE NEUTRON BALANCE

.................................................................................

1 SIX FACTOR FORMULA ........................................................................................................... - 7 Fast Fission Factor - E ................................................................................................................

2 Fast Non-Leakage Probability - -.Li .............................................................................................

3 Resonance Escape Probability - p ..............................................................................................

5 Thermal Non-Leakage Probability - i;h .....................................................................................

7 Thermal Utilization Factor - f ...................................................................................................

7 Reproduction Factor - q .............................................................................................................

9 The Six Factors

...........................................................................................................................

9 FOUR FACTOR FORMULA ....................................................................................................

13 REACTOR CONTROL ..............................................................................................................

13 Moderator-to-Fuel Ratio

...........................................................................................................

14 Effect on Resonance Escape Probability

..................................................................................

15 Effect on Thermal Utilization Factor

.......................................................................................

15 Effect on ken ............................................................................................................................

15 REACTIVITY

............................................................................................................................

16 Excess Reactivity and kexcess ..................................................................................................

18 SHUTDOWN MARGIN ............................................................................................................

20 SDM Demonstration

.................................................................................................................

21 GLOSSARY ................................................................................................................................

22 ... .. .. BWR / REACTOR THEORY / CHAPTER 2 I 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE TABLES AND FIGURES Figure 2-1 Neutron Multiplication

............................................................................................

1 Figure 2-2 Effective Core Size at BOL .....................................................................................

4 Figure 2-3 Effective Core Size at EOL .....................................................................................

4 Figure 2-4 Characteristic Resonance Absorption Cross Section ...............................................

5 Figure 2-5 Neutron Capture and Fission Cross Sections for U-238 ..........................................

5 Figure 2-7 Neutron Reproduction Factor (q) ............................................................................

9 Figure 2-8 Neutron Life Cycle

..................................................................................................

9 Figure 2-6 Plutonium-240 Total Neutron Cross Section ..........................................................

6 Table 2-1 Summary of the Variables Affecting the Six Factor Formula

................................

12 Figure 2-9 Density vs . Temperature

........................................................................................

14 Figure 2- 10 vs . Moderator-to-Fuel Ratio

..........................................................................

14 Figure 2-1 1 p vs . Moderator-to-Fuel Ratio

.............................................................................

15 Figure 2- 12 f vs . Moderator-to-Fuel Ratio

..............................................................................

15 Figure 2- 1 3 vs . Moderator-to-Fuel Ratio ..........................................................................

16 Figure 2-14 bxcess Over Core Life ...........................................................................................

19 -3 . L TaM A 0 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV 3 .. BWR / REACTOR THEORY

/CHAPTER 2 II OBJECTIVES Upon completion of this chapter, the student will be able to perform the following objectives at a minimum proficiency level of 8O%, unless otherwise stated, on an oral or written exam. 1. 2. 3. 4. 5. 6. 7. 8. 9. Describe the neutron life cycle using the followi a. Fast fission factor

b. Fast non-leakage probability factor C. Resonance escape probability factor d. Thermal non-leakage probability factor e. Thermal utilization factor f. Reproduction factor g terms: Define and describe critical, subcritical, and supercritical with respect to the reactor.

Define and describe the effective multiplication factor and discuss its relationship to the state of the reactor.

Define Lxcess. Define shutdown margin.

Define and calculate reactivity.

Explain the relationship between reactivity and effective multiplication factor. Calculate shutdown margin using procedures and given plant parameters.

Evaluate the change in shutdown margin due to changes in plant parameters.

0 2000 GENERAL PHYSICS CORPORATION REV 3 ... BWR / REACTOR THEORY / CHAPTER 2 111 / NEUTRON LIFE CYCLE

-- I KIA # K1.O1 K1.02 K1.03 Kl.04 K1.05 K1.06 - K1.07 K1.08 K1.09 K1.10 K1.ll K1.12 K1.13 I K/A - OBJECTIVE CROSS REFERENCE 3.5* I 3.5 1 I REACTOR THEORY: 292002 NEUTRON LIFE CYCLE 4 XI WA STATEMENT 1.9* 1.9* 2.0* 1.9* ~~ ~ ~~ Describe the neutron life cycle using the fast fission factor. Describe the neutron life cycle using the fast non-leakage probability factor.

1.9* 1.9* 2.1* 2.0* Describe the neutron life cycle using the resonance escape probability factor. Describe the neutron life cycle using the thermal non-leakage probability factor. Describe the neutron life cycle using the thermal utilization factor.

~ 1.9* Describe the neutron life cycle using the reproduction factor.

2.0* Define critical, subcritical, and supercritical with respect to a reactor.

1.9* Define effective multiplication factor and discuss its relationship to the state oj d (sic) reactor. 1.9* Define kcxcess. 3.2 3.2 2.4 1.8* Define shutdown margin.

Define reactivity.

State the relationship between reactivity and effective multiplication factor.

t Calculate shutdown margin using mcedures and given plant parameters.

3.5 5 3.3 6 2.5 7 2.4* 8 IMPORTANCE RO SRO RELATED OBJECTIVE NUMBERS la lb IC Id If 2 3 4 c BWR / REACTOR THEORY

/ CHAPTER 2 iv Q 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE K/A - OBJECTIVE CROSS REFERENCE IMPORTANCE RO SRO REACTOR THEORY: 292002 NEUTRON LIFE CYCLE RELATED OBJECTIVE NUMBERS K/A # K/A STATEMENT K1.14 t Evaluate change in shutdown margin due to changes in plant parameters.

9 -~ ~~ The following objectives, while not cross referenced to specific WAS. ensure mastery of fundamental concepts:

N;A. Note: Importance ratings that are marked with an asterisk (*) or question mark (?) indicate variability in rating responses by reviewers.

An asterisk (*) indicates that the rating spread was very broad. An asterisk (*) can also indicate that more than 15% of the raters felt the knowledge or ability is not required for the ROISRO position at their plant. A question mark (?) indicates that more than 15% of the raters felt that they were not familiar with the knowledge or ability as related to the particular system or design feature.

A dagger (7) indicates that more than 20% of the raters indicated that the level of knowledge or ability required by an SRO is different from the level of knowledge or ability required by an RO. BWR / REACTOR THEORY

/ CHAPTER 2 V Q 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE THIS PAGE INTENTIONALLY BLANK BWR / REACTOR THEORY CHAPTER 2 vi 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE I STEADY STATE f 7 I NEUTRON BALANCE EffOCthM Neutrons In Neutral -----, Muttipticatbrl (Generation

1. l) Fador In th Reactor The neutron population in any given volume depends on the processes that add or remove neutrons from the volume. The time dependent behavior of the neutron population in any reactor at power is given by the mathematical expression:

Neutrons out (Generation

1. 2) + Rate of Rate of Rate of Change of Neutron - Neutron = Neutron Production Population Equation 2-1 When the reactor is in a steady state condition, the rate of neutron production is equal to the rate of neutron removal.

Under these conditions, the rate of change of the neutron population is zero and reactor power will remain constant. Neutrons are produced by fission and removed Several processes determine a neutron's fate. The neutron life cycle represents these various processes and the effects each has on sustaining a steady state condition.

They are discussed in detail later in this chapter. For the purpose of simplification, the following assumptions apply to the neutron life cycle: /1 I by either absorption or leakage from the reactor. All neutrons are born as fast neutrons.

0 Some fast neutrons can be absorbed by he1 and cause fast fission. Some fast neutrons can leak out of the reactor core. 0 Some fast neutrons can be resonantly captured while slowing down.

0 All remaining fast neutrons become thermalized.

p, Some thermal neutrons can leak out of the core. 0 Some thermal neutrons can be absorbed by non-fuel material.

0 Some thermal neutrons can be absorbed by fuel and not cause iission. All remaining thermal neutrons are absorbed by fuel and cause thermal fission.

Figure 2-1 Neutron Multipkation The formula for the effective neutron multiplication factor is: # of neutrons from fission in one generation

1. of neutronsin the previous generation Equation 2-2 kef* = The effective multiplication factor is the product of several factors that affect a neutron during its lifetime. The values of k,, determine whether BWR / REACTOR THEORY /CHAPTER 2 1 of25 Q 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE the neutron population in the core increases, decreases. or remains the same. If the number of neutrons produced by fission in one generation equals the number of neutrons in the previous generation.

k,, = 1. This indicates a steady state condition and defines an exactly critical reactor. If k,, > 1. the number of neutrons produced by fission in one generation is greater than the number of neutrons produced in the previous generation. Subsequently, reactor power increases and the reactor is said to be supercritical.

If ken < 1, the number of neutrons produced by fission in one generation is less than the number of neutrons produced in the previous generation.

When neutron production is less than neutron removal, reactor power will decrease and the reactor is said to be subcritical.

,\ SIX FACTOR FORMULA The six factor formula describes the processes that occur during the neutron life cycle. The starting point in the neutron generation process is the birth of all the fast neutrons from thermal fission events and represents the numerator in the k,, formula. FAST FISSION FACTOR - E In light water reactors, most fissions are caused by thermal neutrons; however, an appreciable number of fast neutrons cause fission in U-235, U-238, and Pu-239. These fissions, known as fast fissions, result in additional fast neutron production above that from thermal fissions.

The fast fission factor (E) accounts for the neutrons produced by fast fission and is given by the equation: fast neutrons produced by ALL fission events fast neutrons produced by THERMAL fission events E= Equation 2-3 Because the fast fission factor represents a net gain in neutron population, the fast fission factor is slightly greater than one, typically between 1.03 and 1.14. In order for fast fission to occur, the neutrons must reach the fuel while they are still fast.

Because the fuel is manufactured as ceramic pellets and the fissions occur within the pellets, the criteria is satisfied. Because of the heavy nucIei within the pellets, the neutrons do not slow down appreciably until reaching the moderator. Once in the moderator, the likelihood of reaching fuel again and causing fast fissions is very small due to the rapid slowing down process. . -_ -- BWR / REACTOR THEORY / CHAPTER 2 2 of25 Q 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE Recall from Chapter 1 that fast fission of U-238 generally requires a neutron with an energy when absorbing a neutron of any energy from thermal to fast. Because delayed neutrons are born with an average energy of 0.5 MeV. fast fission of U-238 is primarily a function of the prompt neutron fraction of the fuel. Even though U-235 comprises a small percentage of the total fuel volume in a commercial reactor core, a large fraction of the fast fissions occurs with U-235 because of its wider fission energy spectrum.

I greater than 1.8 MeV. while U-235 can fission - Parameters affect the fast fission factor by varying the probability of a fast neutron causing fission. The following major parameters affect the fast fission factor:

0 Fuel atomic density - as fuel atomic density decreases, E decreases.

0 Fuel pin diameter - as fuel pin diameter decreases, E decreases. (Fuel pin is the term used for a collection of stacked fuel pellets assembled to produce a fuel bundle.) / encased in metal cladding.

Fuel pins are I \.- Moderation - as the ability of the moderator to slow neutrons down increases, E decreases. Reactor design sets most parameters that have an impact on the value of the fast fission factor during plant operation; thus, the value is not significantly affected by changes in fuel temperature, moderator temperature, or core void faction. Core age is the most significant affect.

As the core is operated, U-235 is depleted, decreasing the fraction of fast fissions from U-235. Even this impact is relatively small and can change the value of E from 1.04 to 1.03 from a new core to a depleted core.

c FAST NON-LEAKAGE PROBABILITY - If As the fast iiciitroii\

produced by fission begin their process ot' don ing down. a possibility exists that a giicii iicwtron will be lost from the core due 10 IC.I~..I+*

The .fast non-kcwkcqy prohiihiljij. ( -fI 1 rL-prc*\ciits the fraction of fast neutrons that dtl not 1c-A out of the core and is given ti! thc cy tr.it it 111 I.I~I ncutrons that

\t.itl to slow down 1.1b1 ricutrtms produced Irtm \I I. fission events I, = Equation 2-4 The fast non-lcah;rgc. prohability represents a net loss in neutron pywldtion and has typical values of 0.90 to 0.90 I hi> means that 90 to 96 percent of fast neutrons rcni;iiii in the core. The ability for a fast neutron to leak out of the reactor is dcpcndcnt upon how far the neutron can travel and its distance from the core boundary.

Therclim.

1, is primarily a function of moderator Jcnsit! and egecrive core six. Effective core size is determined by measuring the average distance (in percent) that the control rods are remold lrorn the physical core. Figure 2-2 represcnis the effective core size of a simplified reactor operating at full power near the beginning of core life (BOL). Figurc 2-3 represents the end of core life (EOL). BWR / REACTOR THEORY

/ CHAPTER 2 3 of25 8 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV 3 ALL NEUTRONS IN ABLE TO LEAK OUT OF THE CORE ' THISAREAARE I CANNOT LEAK OUT OF THE AREA ABOVE THE CONTROL RODS IS CALLED EFFECTIVE CORE SIZE THE CORE li Figure 2-2 Effective Core Size at BOL Evidently, for any steady state operating condition, the fraction of neutrons leaking from the reactor in Figure2-3 is larger than the neutrons not leaking from the reactor in Figure 2-2. The effect of decreased moderator density is to widen the area that the neutrons can leak out of the reactor, whereas, an increased moderator density makes the area narrower. The formation of steam voids in the reactor will decrease the moderator density and increase the probability that a fast neutron can leak from the core. The higher the void fraction, the greater the leakage. ALL NEUTRONS IN THIS AREA ARE ABLE TO LEAK OUT OF THE CORE \ 'NEUTRONS IN THIS AREA A 1 THECORE ' ' THE AREA ABOVE THE CALLED EFFECTIVE CORE SIZE RODS is REPRESENTS AVERAGE CONTROL ROD POSITION Figure 2-3 Effective Core Size at EOL Because the physical core size is so large for the commercial reactor (nearly infinite for neutrons), moderator density has a very minor effect on the value of df. In addition, 4 from BOL to EOL may only change fiom 0.95 to 0.98. Because of this, &is often neglected.

c BWR / REACTOR THEORY

/ CHAPTER 2 4 of 25 8 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV 3 RESONANCE ESCAPE I PROBABILITY - p c/ All nuclei have some probability of absorbing a neutron as indicated by the microscopic cross section for absorption (OJ. The microscopic cross section for absorption is not a constant value. but it is dependent on the energy level of the neutron. In general, the cross section for absorption increases as the neutron energy level decreases. However, certain nuclei (U-238 and Pu-240 in particular) show an extremely high absorption cross section for neutrons at specific energy levels.

At certain neutron energy levels, the cross section can be as much as 1,000 times the cross section for a neutron of a slightly higher or lower energy level (Figure 2-4). INTERMEDIATE , FAST (EPITHERMAL)

I RESONANCE

I I I I I , I II I1 I I 111 IO" io" 1 o 10 10' 103 10' IO' lo6 10' ev io4 10" 10' ioJ lo4 10' 10.' 10' 1.0 10 MeV NEUTRON ENERGY Figure 2-4 Characteristic Resonance Absorption Cross Section These high values are called resonance absorption peaks.

These specific energy levels represent vacant energy sites for a nucleus. After a neutron is born from fission and begins to slow to thermal energy levels. it passes through the resonance regions of the core materials.

When the neutron is at a resonance energy level. the probability of being absorbed by the non-fuel material is very high. This process is hnown as resonance capture. Those neutrons resonantly captured are lost to the fission process for the generation of neutrons.

Figure2-5 shows both the neutron capture and fission cross sections for U-238. Note that the capture cross section (the n, y reaction) shows a number of resonance peaks in the slowing down or intermediate region (between 1 eV and IO5 eV). The largest of these resonance peaks is about 7,000 barns and it occurs for an incident neutron energy of about 6.7eV. Note that fission of U-238 is not probable unless the incident neutron energy is in the 1 MeV (IO6 eV) range (for further discussion, see Chapter 1, Neutrons).

Figure 2-5 Neutron Capture and Fission Cross Sections for U-238 BWR / REACTOR THEORY / CHAPTER 2 5 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 L / NEUTRON LIFE CYCLE Figure 2-6 shows the total cross section for Pu-240. Although labeled "total", it largely represents capture (the n, y reaction) below 1 O6 eV and fission above 1 O6 eV ( 1 MeV) (refer to Chapter 1. Neutrons). Note that Pu-240 has a giant resonance peak of = 1.2 x 1 O5 barns at about 1 eV of incident neutron energy. Other resonance peaks are considerably smaller and occur in the slowing down or intermediate range as we1 I. SLOW INTERYEUATE FAST (THERMAL) (EPt"HERYAL) 1OMlDOO- I I I I amoN1w.?40 TOTM cmss 0 10 - I 10 - I I I I 01-' 1' " IO? 10' io io id id io' id id io' sv io+ io' io' io' ioA io' IO' io' 10 io ysv MFFERLHTUL ENERGY Figure 2-6 Pluloniunt-240 Total Neutron Cross Section The resonance escape probabiiiw (p) is the fraction of neutrons that are not absorbed while slowing to thermal energy. fast neutrons that become thermal fast neutrons that start to slow down P= Equation 2-5 The resonance escape probability represents a net loss in neutron population and has typical values of 0.74 to 0.80. Several factors affect the value of the resonance escape probability, such as the moderator-to-fuel ratio, fuel temperature, core age, and he1 enrichment.

The resonance escrtpc probability also varies with changes in tucl temperature and core age.

Increasing the temperature of the fuel causes increased re:Son;iiicc absorption, decreasing the resonance escapc prohrtbility.

J During thc lilk 01' the core, some U-238 transforms inio 1'11-2-10.

as shown in the followi ng L'LILI;~I ion. The production 01 I'u-240. as U-238 slowly depletes mcr cow lilt. rcsults in an even higher resonancc ahsorkr ( I'u-240) replacing a high one (U-238). I hcrctiwc.

there is an increase in resonance cclpturc tncr core life.

As a result, the resonance cscapc prcihability over core life will decrease.

A t!,pical value for the resonance escape probabilit?, (p) is about 0.75. P Increasing fucl cnrichrnent (concentration of U-235 atoms) will haix a minor effect (increase) on the resonancc cscapc probability.

This is due to the decrease in U-238 concentration which decreases the amount of neutrons absorbed in U-238. Resonance absorption is also affected by the time it takes for the neutron IO slow down to thermal energies. This timc is inversely proportional to the density of the moderator and is directly proportional to f he t-oid coefficient. Resonance absorption is discussed in more detail in Chapter 4. Improved moderation causes the neutrons to slow down faster.

Therefore, a given neutron spends less time in the resonant regions, decreasing the probability of resonance absorption.

This causes the value of p to increase.

BWR / REACTOR THEORY / CHAPTER 2 6 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE THERMAL NON-LEAKAGE PROBABILITY - & As thermal neutrons begin the diffusion process. a possibility exists that some of the neutrons will be lost to core leakage.

The thermal non-leakup probabiliQ

(&) represents the probability that a thermal neutron will not leak out of the core and is given by the following equation: thermal neutrons absorbed in the core fast neutrons that become thermal

= Equation 2-7 The thermal non-leakage probability is impacted by the same parameters (effective core size and moderator density) as the fast non-leakage probability. The effect of these parameters is less because the distance that a neutron travels in the thermal energy range is much less than that of a fast neutron.

As with the fast non-leakage probability, thermal non-leakage probability decreases with increased void coefficient.

The thermal non-leakage probability represents a net loss in the neutron population and has a typical value of 0.98. As with the fast non- leakage probability, this leakage term is often neglected due to the relative infinite size of the reactor. A factor so close to 1 .O does not change the value of very much.

,? THERMAL UTILIZATION FACTOR - f All materials in the reactor absorb neutrons to some extent. By carefully selecting the materials that go into the reactor, control of neutron absorption is accomplished and non-fuel absorption is minimized.

The thermal iiti/izution .factor (f) is the ratio of the number of thenpal neutrons absorbed in the fuel to the number of thermal neutrons absorbed in the core. The term core includes the fuel, moderator, fuel cladding, structural members, control rods, etc. thermal neutrons absorbed in fuel thermal neutrons absorbed in core f= Equation 2-8 The thermal utilization factor represents a net loss in neutron population and has a typical value of 0.70 to 0.88. f- BWR / REACTOR THEORY

/CHAPTER 2 7 of25 Q 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE For simplification, the thermal neutron absorbers in core materials, other than the fuel or moderator, are combined in one term designated as "other" or "0". Future chapters will detail the other terms: the other term can be examined further by delineating it into control rods. poisons, and others. In this discussion, assume the neutron flux for the fuel, moderator.

and others is equal. The formula for thermal utilization factor is as follows: Dividing by (Vf,,l$) gives: P fuel Where: V = volume 6 = thermalneutron flux c, = macroscopic cross section for absorption Equation 2-9 The simplified equation is: Where: 2, = macroscopic cross section for absorption fuel = reactor fuel mod = moderator o = other thermal neutron absorbers in core Equation 2-I 0 Withdrawing control rods decreases the macroscopic cross section for absorption of the "other" materials. thereby allowing the operator to increase reactor power. Inserting control rods increases the macroscopic cross section for absorption of the "other" materials, causing the thermal utilization factor to decrease.

The decrease in the thermal utilization factor causes reactor power to decrease. The thermal utilization factor is the main factor that the reactor operator changes to control k,, ) Over core life, several variables affect the thermal utilization factor.

A significant amount of he1 is added for extended operations (typically 18 months or longer). Increasing the enrichment (enrichment is the ratio of U-235 atoms to the total number of uranium atoms) of the fuel causes an increase in the thermal utilization factor by increasing the ratio of U-235 atoms to U-238 atoms. This is because the thermal neutron macroscopic cross section for absorption for U-235 is greater than the cross section for U-238. \ Over core life, fuel enrichment decreases with burnup of the U-235 (decreasing the ratio of U-235 atoms to the total number of uranium atoms). This causes a decrease in the thermal utilization factor. This is because the thermal neutron macroscopic cross section for absorption of U-238 is much less than that of U-235. -- As the fuel decreases, the moderator-to-hel ratio increases and neutron absorption in the moderator increases. The number of neutrons available to be absorbed in the fuel decreases, thereby decreasing the thermal utilization factor.

As the core ages, gadolinium (burnable poison installed for reactivity control) depletes and Pu-239 builds up as the result of U-238 neutron captures. Combined, these factors tend to offset the U-235 depletion and can result in a slight increase from BOL to EOL. Typical values change from 0.76 at BOL to 0.78 at EOL. - BWR / REACTOR THEORY /CHAPTER 2 8 of25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE REPRODUCTION FACTOR - Y' -.. .' I , The reproduction

$mor (q) represents the number of fast neutrons produced from fission compared to the number of thermal neutrons absorbed in the fuel. as shown in Figure 2-7. Fast Neutrons Out Thermal - Neutrons In Figure 2- 7 Neutron Reproduction Factor (q) In equation form. q becomes: fast neutrons produced by thermal fission events thermal neutrons absorbed in the fuel rl= Equation 2-11 To determine this value, the macroscopic cross section for fission of the fuel is multiplied by the ' neutron yield per fission (v). Taking into account that not all thermal neutrons absorbed in the fuel result in fission, this term is divided by the sum of the macroscopic cross sections for absorption of all fuels present in the core.

---

, Where: = average number of neutrons produced for each neutron absorbed cf = macroscopic cross section for fission xa = macroscopic cross section for absorption Equation 2-12 .\ The reproduction factor represents a net gain in neutron population and has typical values of 1.74 to 2.00. The value varies with fuel enrichment and core age. Increasing the enrichment causes an increase in the reproduction factor (the macroscopic cross section for fission of U-235 is larger). As the core ages, Pu-239 is produced.

Although the neutron yield per fission for Pu-239 is slightly higher than for U-235. the production of Pu-239 lags the depletion of U-235. The combination of these factors leads to a slight decrease in the reproduction factor over core life. THE SIX FACTORS Figure 2-8 Neutron Life Cycle Using Figure 2-8, assume that the neutron life cycle begins with 1800 fast neutrons. Recall, that these fast neutrons are born from thermal fission of U-235 fuel. B WR / REACTOR THEORY / CHAPTER 2 9 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 1 NEUTRON LIFE CYCLE Of these neutrons, some will cause fast fission in U-235, U-238, and Pu-239, which will produce additional fast neutrons. This equation represents the fast fission factor (E): fast neutrons produced by ALL fission events fast neutrons produced by THERMAL fission events E= Equation 2-I3 If the number of fast neutrons increases from 1800 to 1854, the fast fission factor can be determined.

Then: - 1.03 1854 1800 &=-- Example 2-1 Thus, in this example, the fast fission factor (E) is 1.03. The I854 fast neutrons exist to continue through the neutron life cycle. Some of the fast neutrons will be lost due to fast leakage. The fast non-leakage probability (if) represents the fraction of fast neutrons that do - not leak out of the core and is given by: fast neutrons that start to slow down fast neutrons produced from ALL fission events -t;= Equation 2-14 If 74 fast neutrons leak out of the core, the fast non-leakage factor can be determined.

Then: 1, =-- - 0.96 1854 Example 2-2 In this example, 1780 fast neutrons remain in the core and begin to slow down.

Therefore, the 1, is 0.96. As the remaining neutrons begin to slow down, they pass though the resonance region and are subject to resonance capture. The resonance escape probability defines the probability that a given neutron will escape capture and is given by : fast neutrons that become thermal Y- fast neutrons that start to slow down Equation 2-15 If it is determined that 427 neutrons are absorbed in the resonance peak regions, then the resonance escape probability can be calculated.

Then: p=-= '353 0.76 I 1780 I Example 2-3 Therefore, 1353 neutrons reach thermal energy and p = 0.76. A fraction of the thermal neutrons will be lost to thermal leakage. The fraction of neutrons that are not lost is given by the thermal non-leakage probability

(-t;h) and is given by: BWR / REACTOR THEORY I' CHAPTER 2 10 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE thermal neutrons absorbed in the core fast neutrons that become thermal lth = Then: Then: _I 1326 1353 l,h =-= 0.98 Equation 2-1 6 If 27 thermal neutrons leak out of the core, 1326 thermal neutrons remain to be absorbed in the core' (fuel and non-fuel materials).

Example 2-4 fast neutrons produced by thermal fission events thermal neutrons absorbed in the fuel rl= Equation 2-18 There are 1000 thermal neutrons available for absorption into the U-235 fuel. Because of these absorptions, fast neutrons are born from fission. The fission process produces 1800 fast neutrons.

Therefore, in this example, 41, equals 0.98. Example 2-6 The next factor to be determined in this neutron life cycle is the thermal utilization factor (f). It is written mathematically: thermal neutrons absorbed in fuel themal neutrons absorbed in the core f= Equation 2-1 7 If 326 neutrons are absorbed into non-fuel atoms in the core, 1000 neutrons remain to be absorbed into the fuel. C Then: Example 2-5 The thermal utilization factor (f = 0.754) denotes the ratio of the thermal neutron absorbed in the fuel to those absorbed in the core. The last factor to be considered is the reproduction factor (q). This equation shows the reproduction factor. In this example, the neutron reproduction factor (11) equals 1.80. Note that the number of neutrons produced by fission in this generation equals the number of neutrons produced in the previous generation.

By definition, km is equal to one and the reactor is exactly critical. The effective multiplication factor (kff) equals the product of the six factors and is independent of neutron sources other than fission. Equation 2-1 9 BWR / REACTOR THEORY / CHAPTER 2 11 of25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE Using the six factors as determined from Figure 2-8, bn becomes: + 4 kern = Reactor Pressure +e+++ Moderator Temperature

+*++e 854 1780 1353 1326 800 1854 1780 1353 - X- x-x- + + 4 000 1801 326 100( X- - 1.03 x 0.96 x 0.76 x 0.98 x 0.754 x 1 .so k,, = 1 .O (The reactor is critical.)

Core Age ++*+++ Fuel Temperature Void Content +++++ Fission Product Poisons Example 2- 7 The effective multiplication factor (ken) is essentially a measure of the probability that one illustrated by the six factor formula, core size and materials affect this probability.

However, there is no affect by the introduction of non- fission neutrons.

fission event will cause another fission.

AS -4 Table 2-1 Summary of the Variables Affecting the Sir Factor Formula I I VARIABLE I E 14 P l&l f 11.11 NOTES: Denotes major effect + Denotes minor effect All changes listed above assume that no other variable is affected by the change indicated. Since the core size of a commercial nuclear reactor is large, changes to the non-leakage probability factors (4 and l*) are minor. Other parameters that can have an effect on the neutron life cycle include: core design parameters, change in concentration of other fissile and fissionable materials, and concentration of burnable poisons. These parameters have not been included as they are not affected by operator action.

BWR / REACTOR THEORY

/ CHAPTER 2 12 of25 0 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV3 I FOUR FACTOR FORMULA I L 8 - ( Because the size of a commercial reactor is very large relative to neutron travel, the core is considered to have infinite volume. Considered insignificant. the non-leakage factors

(-& and 4th) may be omitted from the six factor formula.

The resulting equation, called the .four fucfor .formula. describes the infinite multiplication factor (k). k,=Epfq Equation 2-20 I REACTORCONTROL I ~ In order to control reactor power, the operator must be able to control the thermal neutron population. Controlling or varying the values of the factors that affect neutron multiplication controls the thermal neutron population.

As previously stated, the non-leakage factors are insignificant. This leaves the four factor formula for reactor control. Although the fast fission factor (E) and reproduction factor (q) are important for neutron production, both have values that are primarily a function of reactor design and remain essentially constant in the temperature range of an operating reactor. Reactor parameters like plutonium concentration, fuel enrichment, and poisons have negligible effects on the value of E. Uranium and plutonium have both thermal and fast-fission isotopes.

In a uranium fbeled reactor, as uranium is burned, plutonium is produced.

These changes tend to counterbalance each other, and the value of E remains fairly constant. Poisons absorb neutrons in the thermal and epithemal ranges. Therefore, they have no effect on the value of E. The nominal value of E is slightly greater than one. Thus, only the resonance escape probability (p) and the thermal utilization factor (f) have significant changing values and are major factors affecting the control of reactor power. BWR / REACTOR THEORY /CHAPTER 2 13 of25 Q 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE MODE RATOR-TO-F U E L RAT1 0 The ratio of moderator atoms to fuel atoms has a pronounced effect on both the resonance escape probability and the thermal utilization factor. Understanding these effects requires a short discussion on the moderator-to-fuel ratio. This ratio can be modified by changing the fuel pin lattice spacing or the moderator density. The spacing of fuel pins in the fuel bundles is set by reactor design and is not controlled by the operator. However, moderator density in the BWR core is affected by moderator temperature changes, which the operator can directly control.

AP I Am AT, = AT2 432 > 4m Figure 2-9 Density vs. Temperature Figure 2-9 shows that water density decreases as temperature increases.

Also, the slope of the graph is much steeper at higher temperatures.

This observation is very important.

It means that for the same change in temperature, the change in water density is greater at higher temperatures.

The slope of moderator-to-he1 ratio (NmodNfuc~)

versus temperature closely approximates the slope of water density versus temperature.

Thus, as moderator temperature increases, NAhl decreases and the change is larger at a higher temperature.

The reactor is designed to operate at relatively constant temperatures and pressures. Once the reactor is at operating pressure, moderator temperature changes are relatively small. However, moderator temperature changes occur in the reactor during reactor startups and heatups to operating temperature. Moderator-to-fuel ratio at a given temperature is not a parameter directly controlled by the operator. The value of moderator-to-fuel ratio is designed to be less than ideal to maximize utilization of all neutrons. This condition leads to a negative temperature coefficient and is called undermoderated.

As moderator temperature increases and density decreases, a constant water level in the vessel will yield a further drop in the moderator-to-fuel ratio and an insertion of negative reactivity, as shown in Figure 2-1 0. -l J UNDER OMR - MODERATION 1'4- MODERATION-EGlNNlNG of CORE LIFE END OF CORE LIFE (COLD) (COLD)=24 1 .- 1 MOOERATOR TO FUEL RAM (Ha0 Wu) Figure 2-10 keg vs. Moderator-to-Fuel Ratio In a reactor where the water-to-he1 ratio is considerably high, an overmoderated condition exists. Here, the decrease in neutron absorption in water with a density decrease overshadows neutron losses from leakage, resonances, and neutron absorption in hydrogen and oxygen nuclei. Since more neutrons are available in the fuel, this leads to positive reactivity.

BWR / REACTOR THEORY /CHAPTER 2 14 of 25 8 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV 3 Later in core life, the large reduction in fuel molecules and the decrease in moderator density during a plant heatup can lead to a positive -& insertion. This event occurs at some BWRs and is seen at temperatures around 200" to 220°F. d- EFFECT ON RESONANCE ESCAPE PROBABILITY Figure 2-1 1 depicts a graph of resonance escape probability (p) versus moderator-to-fuel ratio (and temperature).

As moderator temperature increases, moderator density decreases and moderator-to-fuel ratio decreases.

A fast neutron travels farther to undergo the same number of collisions to reach thermal equilibrium. The chance of a neutron being captured at resonance energies increases. Therefore, resonance escape probability decreases as moderator temperature increases. The slope of the graph in Figure 2-1 1 is steeper at a lower moderator-to-he1 ratio because of the larger density change of the moderator at higher temperature.

r AS NwdNru~~ GETS LARGER. THE FRACTION OF NEUTRONS THAT ESCAPE RESONANCE CAPTURE INCREASES (p INCREASES) 4 1 ._,-.-*- P MOOERATOR TEMPERATURE Figure 2-1 I p vs. Moderator-to-Fuel Ratio EFFECT ON THERMAL UTI LlZATlO N FACTOR - Figure 2-12 dcpicts a graph of the thermal utilization factor ( 1') \usus moderator-to-fuel ratio. As the dciisii: ot. the moderator decreases.

less modcrmtr atcwilr arc available for thermal neutron absorpioii in non-fuel materials.

This is significant in ;I crmiiiicrcial reactor because thermal ncutrcui llu\ tends to be higher in the moderator

ind hwcr in the fuel pins. The effect is more proiicwiic.cJ at higher moderator ternpcraturc>

tx-c.~uw of the greater change in density of tlic iiIoJcraitvr at higher temperatures.

AS NU ..n+ , CC TS LARGER. THE FRACTION OF DECW ASLS I' DECREASES)

NEUTH )NI> THAT ARE ABSORBED INTHE FUEL u; 0 --- 5 06 Figure 2-1 2 f IX Alderator-to-Fuel Ratio EFFECT ON ken ~~~ ~~ Multiplying the value of each term of the six factor formula. uhich exists at a given value of Nmod/Nfue,.

results in the k~ curve shoun in Figure 2-1 3. Maintaining at 1 .O is necessa? to maintain constant reactor power. Obsen ing the shape of the cuwe. k=r increases rapidl). through 1.0. reaches a maximum. and then decreases slowly past 1 .O again. BWR ! REACTOR THEORY /CHAPTER 2 15 of25 0 2000 GENERAL PHYSICS CORPORATION REV 3 C / NEUTRON LIFE CYCLE Since bm governs the reactor power, analyzing the curve predicts the response of the reactor.

When a reactor operates in the region to the left of maximum kern, a temperature increase will lower bff, decreasing neutron population and lowering reactor power.

A new lower reactor power level will cool the fuel and moderator.

A new equilibrium is attained with back at 1.0. Considered the stable region of the curve, it is called the undermoderated region. If a reactor is operated to the right of maximum btf, a moderator temperature increase would increase kern. Reactor power would increase which would raise moderator temperature further and increase bn even more. This is called the overmoderated region. The ovennoderated region is less stable and is undesirable for commercial reactors in the United States. 14 12 1 .o Ilr 0.8 0.6 0.4 Figure 2-13 kc* vs. Moderator-to-Fuel Ratio 1 REACTIVITY Reactivity is the measure of the departure of a reactor from criticality. Reactivity is defined as the fractional change in neutron population per generation and is indicated by the Greek letter rho (p). The fractional change in neutron population per generation (reactivity) can be shown by the equation given below.

Equation 2-21 Calculate the reactivity level of a core with a k,,of 0.985. Example 2-8 The following notational changes simplify the discussion of reactivity.

Equation 2-22 BWR / REACTOR THEORY / CHAPTER 2 16 of25 6 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE In addition to the hWk unit for reactivity, the fractional change in neutron population can be expressed in terms of % Mk. The % Auk unit can be obtained as follows: i L A shutdown reactor has a core reactivity of -0.0038 Ak/k. Calculate the core reactivity value in %Ak/k. J ,($) x 100% = p(%Ak i k) 1-P Equation 2-24 A shutdown reactor has a core reactivity of -0.0028 Akk. Calculate the core k,, . Equation 2-23 Example 2-10 Example 2-1 I further illustrates these reactivity units. I From the previous example, the reactivity level of a core is -0.01 52 Akk. Calculate the core reactivity value in % Ak/k. A control rod withdrawal results in the k,, of a reactor changing from 0.97 to 0.975. How much reactivity is added to the core by the control rod withdrawal? Example 2-11 BWR / REACTOR THEORY /CHAPTER 2 I7 of25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE Reactivity is a convenient term to use when discussing deviations from criticality.

For any power, if the reactor is critical (ke, = 1). the reactivity associated with the reactor is zero. For a supercritical reactor, reactivity is a positive value, and for a subcritical reactor. reactivity is a negative value. Rearranging the and reactivity equation solves for reactivity:

1 k =- ClT I-P 1 p=]-- ke* Equation 2-25 If k,, equation, reactivity is equal to zero. is equal to 1, substituting into the I 1 p = 1 --= 1 --=o p=O k, 1 If k,, is greater than 1, substituting into the equation, reactivity is a positive value.

1 1 p = 1 - - = 1 - - = 0.000999 ken 1.001 p is positive If ken is less than 1, substituting into the equation, reactivity is a negative value. 1 - = -0.001 p = 1 --- 1 kefi 0.999 p is negative Example 2-13 EXCESS REACTIVITY AND kexcess Excess ~-e~~ii\-i~~* (pc,ccss) is the reactivity 3 associated uith (tic cscess fuel that is added to the core be! orid thc minimum amount necessary to achiet.e criticalit!

at BOL. The critical core always has Lcll -7 I. The "excess" reactivity canccls out u itli coiitrot rods and/or recirculation flow ratcs. I litring [tic reactor core's lifetime.

core rcxtit it! dccrcascs for the following reasons: 1 tUcl huriiup. 2) fission product poison buildup. and 3 rcsonant absorber buildup (Pu-240).

'I'hcrclivc.

thc reactor has an excess potential rc;icti\ it! initially built in to compensatc Iiv ilic~c dccrcases.

In addition.

excess rcacti\it>

i3 ddcd to address the short term nrgati\c rcxtit it! inputs that occur from reactor stanup to lull power operation, which is related primaril\ iiiodcrator density decreases.

The definition 01' CY~~.SS multiplication .factor (bxcess) is thc rrriiwril hy which the total installed bfi exceeds 1 .O. It is mathematically expressed as : I - L'.,, - kcfT - 1 Therefore:

~ ~- BWR REACTOR THEORY / CHAPTER 2 18 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 I NEUTRON LIFE CYCLE I A reactor is in a refueling outage.

Fuel has been added to the core, raising the total core k,, value to 1.4. Calculate the value of core excess reactivity (pexcess). ~~ ~ Example 2-14 ~,,,,, and excess reactivity are generally defined for specific conditions. Commonly used conditions are: ' 3 0 Cold, xenon-free, unrodded 0 Hot, xenon-free, unrodded 0 Hot, rated power, equilibrium fission product poisons (xenon and samarium)

The value of ~,,,,, will vary over core life (Figure2-14).

At the beginning of core life (point A to point B). bxce,s decreases due to samarium and xenon (fission product poisons) buildup in the reactor. Excess multiplication factor (kexcess) increases to a maximum value (point C) because of depleting burnable poisons that are added to the core during refueling.

Excess multiplication factor (kxcess) then decreases due to he1 burnout, until bxcess is exhausted (point D). At this time, the excess reactivity is also zero and coastdown begins. The reactor is shut down for refueling once the target fuel burnup is reached. This occurs at some point past point D. C \ WITH BURNABLE POISONS 01 CORE AGE Figure 2-14 k,,, Over Core Life \ T BWR / REACTOR THEORY / CHAPTER 2 19 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE I SHUTDOWNMARGIN I -- Technical specifications define the shutdown margin (SDM) as the amount of reactivity by which a xenon-free, cold (68°F) reactor would be subcritical if all but the highest worth control rod were fully inserted. The highest worth control rod is assumed to be fully withdrawn.

The shutdown margin for a subcritical reactor can be calculated by using the following equation:

Equation 2-2 7 Note that this equation is different from the reactivity equation; the terms in the numerator are reversed.

Any parameter that varies core reactivity causes the shutdown margin to change (e.g., control rod density changes, moderator density changes, poison concentration changes, etc.). If the core reactivity becomes less negative, the shutdown margin will decrease.

\A Calculate the shutdown margin of a shutdown reactor with a core reactivity value of -0.0045 Auk. Example 2-1 5 Core design and existing conditions determine the amount of reactivity by which a reactor is actually shut down. The following parameters or design features affect shutdown reactivity conditions: - 0 Moderator temperature - An increase inserts negative reactivity, increasing the shutdown margin. 0 Fuel temperature - An increase inserts negative reactivity, increasing the shutdown margin. 0 Control rod position - A rod insertion adds negative reactivity, increasing the shutdown margin. 0 Xenon concentration - An increase adds negative reactivity, increasing the shutdown margin. BWR / REACTOR THEORY /CHAPTER 2 20 of 25 0 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV 3 0 Number of fuel assemblies in the core - A Typically, SDM determination is required for removal of fuel assemblies adds negative specific reactor core conditions and/or rod reactivity. increasing the shutdown margin control inoperability as specified by plant during refueling.

technical specifications.

7 0 Exposurehumup of fuel assemblies in the core - An increase in exposure or burnup adds negative reactivity.

increasing the shutdown margin.

SDM DEMONSTRATION SDM is demonstrated by withdrawing control rods to achieve criticality with a stable reactor period. Using the formula listed below, the SDM is empirically derived by adjusting the following factors: SDM = (a - b+c -d)100% = %Ak / k Where: a = worth of all withdrawn control rods (The reactivity that would be added if all withdrawn rods are inserted.)

b = worth of most reactive control rod (Assumes the most reactive control rod is fully withdrawn.)

\ 9 c = Moderator temperature correction factor (The reactivity that would be added by the change in moderator temperature to 68OF.) d = Reactor period correction factor (A measure of the current state of the reactor regarding its departure from criticality.)

Equation 2-28 Chapters 5, 6, 7, and 8 discuss in detail the factors affecting SDM. ,n BWR / REACTOR THEORY / CHAPTER 2 21 of25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE Critical Effective Multiplication Factor (kern) Excess Multiplication Factor (kexcess) Excess Reactivity (pexccss)

Fast Fission Factor (E) Fast Non-Leakage Probability (1;) Infinite Multiplication Factor (k) Reactivity (p) Reproduction Factor (q) Resonance Escape Probability (p) GLOSSARY The condition of the reactor where the number of neutrons produced by fission in one generation equals the number of neutrons produced by fission in the previous generation (kern= 1) (p = 0). The factor by which the number of neutrons produced by fission in one generation must be multiplied to determine the number of neutrons produced by fission in the next generation. The amount by which the total installed core exceeds 1 .O. The reactivity added to the core over and above that needed to achieve criticality. Excess reactivity is built into a reactor to compensate for fuel burnup, accumulation of fission product poisons, resonant absorber buildup, and increased temperature from shutdown to power operations. The ratio of fast neutrons produced from all fission events divided by fast neutrons produced by thermal fission events. The ratio of the number of fast neutrons that start to slow down divided by the number of fast neutrons produced from all fissions.

The number of neutrons produced from fission in one generation divided by the number of neutrons produced from fission in the previous generation in a reactor of infinite size (i.e., neutron leakage does &occur). The fractional change in neutron population per generation, or the measure of the departure of a reactor from criticality. Reactivity is zero when the reactor is exactly critical. If positive reactivity is added, reactor power will increase. If negative reactivity is added, reactor power will decrease.

The ratio of fast neutrons produced by thermal fission events divided by the number of thermal neutrons absorbed in the fuel. The ratio of fast neutrons that become thermal divided by the number of fast neutrons that start to slow down.

BWR / REACTOR THEORY /CHAPTER 2 22 of25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE Shutdown Margin (SDM) Subcritical Supercritical Thermal Non-Leakage Factor (&) Thermal Uti1 ization Probability (f) GLOSSARY The amount of reactivity by which a xenon-free, cold (68°F) reactor is or would be subcritical if all but the highest worth control rods were fully inserted. The highest worth control rod is assumed to be hlly withdrawn. The condition in which the number of neutrons produced by fission in one generation is less than the number of neutrons produced by fission in the previous generation (bn < 1) (negative p). The condition in which the number of neutrons produced by fission in one generation is greater than the number of neutrons produced by fission in the previous generation (ken > 1) (positive p). The ratio of the number of thermal neutrons absorbed in the core divided by the number of fast neutrons that become thermal. The ratio of the number of thermal neutrons absorbed in fuel divided by the number of thermal neutrons absorbed in the core.

BWR / REACTOR THEORY /CHAPTER 2 23 of25 0 2000 GENERAL PHYSICS CORPORATION 1 NEUTRON LIFE CYCLE REV 3 I Calculate the reactivity level of a core with a k,, of 0.985. \ EXAMPLE ANSWERS k,, - ' k,ff P= 0.985 - 1 0.985 P= P = -0.0152 Ak / k From the previous example, the reactivity level of a core is -0.01 52 M. Calculate the core reactivity value in YO Akk. (- O.O152~k/

k)x 100% = -1.52%& / k &ample 2-9 A control rod withdrawal results in the k,,T of a reactor changing from 0.97 to 0.975. How much reactivity is added to the core by the control rod withdrawal?

0*975 - = -0.0256~k

/ k - pz - 0.975 0.97 - 1 = -0.0309Ak

/ k = 0.97 *P = Pz -P1 Ap = -0.0256Ak/k-(-0.0309Ak/k)

Ap = 0.0053Ak/k or Ap = 0.53%Ak/k Example 2-1 I A shutdown reactor has a core reactivity of -0.0038 Akk. Calculate the core reactivity value in %Ak/k. (-0.0038~k

/ k) x 100% = - 0.38%Ak / k Example 2- IO Example 2-22 A shutdown reactor has a core reactivity of -0.0028 Ak/k. Calculate the core keK . 1 k,, =- I -P 1 1 - (-0.0028) k,, = k,, = 0.9972 --_ - /-- .f BWR / REACTOR THEORY

/ CHAPTER 2 24 of 25 0 2000 GENERAL PHYSICS CORPORATION REV 3 / NEUTRON LIFE CYCLE I EXAMPLE ANSWERS - ~~ ~~ ~~ If k,, is equal to 1, substituting into the equation, reactivity is equal to zero. SDM = 0.0045Akk I p=O If ke, is greater than I, substituting into the equation, reactivity is a positive value. 1 I - - = 0.000999 I p=]--= kCff 1.001 p is positive If k,, is less than 1, substituting into the equation, reactivity is a negative value. 1 1 -- = -0.001 p=l--= 1 kCff 0.999 p is negative ~ ~- Example 2-13 A reactor is in a refueling outage. Fuel has been added to the core. raising the total core k,, value to 1.4. Calculate the value of core excess reactivity (pexcess).

kCxCCSS = 1.4 - I = 0.4 0.4 1.4 -- - k excess kCfT Pcxc,, - - - P CXEeSS = 0.286 Akk PWCS = 28.6%Ak/k Barnpie 2-14 Calculate the shutdown margin of a shutdown reactor with a core reactivity value of -0.0045 Ak/k. 1 1 k =-= 1 - p 1 - (-0.0045) eff k, =0.9955 1 - 0.9955 SDM = 0.9955 Example 2-15 BWR / REACTOR THEORY

/ CHAPTER 2 25 of25 0 2000 GENERAL PHYSICS CORPORATION

/NEUTRON LIFE CYCLE REV 3 Question Number: 37 (Missed by all three candidates) Facility Regrade Request:

Change the correct answer to "c" Justification:

The question provides that Reactor Building Closed Cooling Water System (RBCCW) is backing up Turbine Building Closed Cooling Water System (TBCCW), and Reactor Water Cleanup (RWCU) is dumping 60 gpm to the main condenser.

It also provides that Control Rod Drive System (CRD) is in service, which provides a normal flow of approximately 60 gpm to the reactor vessel. With RWCU dump flow out of the reactor compensating for CRD flow into the reactor, reactor cavitykpent fuel pool level will be stable. The question then states that both RBCCW pumps trip. The answer key indicates that the operational implication of this would be that Reactor Cavity and Fuel Pool water level will begin to lower, which is answer "d. The justification for this on the answer key states that "The CRD pumps will trip after a loss of RBCCW." If this were the case and the running RWCU pump remains running, then reactor cavitykpent fuel pool level would lower.

However, there is no direct trip of the CRD pump due to a loss of RBCCW flow. The PBAPS Initial Licensed Operator Training lesson plan for Control Rod Drive Hydraulic System, PLOT-5003A states on page 18 of 24, under "Interlocks", that the pump will trip on low suction pressure and various electrical malfunctions.

This is also supported by the Annunciator Response Card for CRD WATER PUMP TRIP (ARC-21 1, F-1 and G-1) that lists only "Low suction pressure" and "Motor overcurrent" as the automatic trips of the CRD pumps. The lesson plan also states (on page 20 of 24) that "A loss of TBCCW and RBCCW will cause the CRD pump to overheat." Therefore, the most that can be said for the CRD pump on a loss of RBCCW and TBCCW is that it will not automatically trip, but it may trip due to overcurrent as a result of overheating.

Since RBCCW also cools the RWCU pump motor coolers (see Design Basis Documents P-S-33 for RBCCW and P-S-36 for RWCU) a loss of RBCCW yiJ result in an automatic trip of the RWCU pump due to high temperature in the RWCU pump motor windings at a setpoint of 149 deg. F. This is supported by ARC-21 5, A-2 and 6-2, which are provided.

In addition, the RWCU System Manager at PBAPS, Luis Feliu (71 7-456-3634) indicated that this trip would occur "fairly soon" after a loss of RBCCW at rated conditions.

He also indicated that with the reactor shutdown and cooled down to a temperature typically seen during a refueling outage, the high temperature trip setpoint may take longer to reach due to the absence of heat conduction input from the system, but would still reach the high temperature trip setpoint due to the heat generated due to the motor winding current. This was also confirmed by the alternate RWCU System Manager at PBAPS, who was the previous RWCU System Manager, as well as engineering personnel at LGS, which has similar RWCU pumps. The System Managers also indicated even with the RPV flooded up to normal level for refueling operations, no dump flow would be expected after RWCU pump trip, due to the lack of RPV pressure and the high headloss of the circuitous RWCU dump flowpath (RPV pressure will be approximately 0 psig since the reactor is in Mode 5 with Core Shuffle Part 1 in progress). When the author of this question was asked why RWCU was assumed to remain in service, he responded that he overlooked the high motor winding trip for the RWCU pumps. Since RWCU will trip, and CRD may or may not trip, reactor cavity and spent fuel pool level will not lower, therefore, answer "d is not correct. Spent fuel bundles are covered with a loose coating of corrosion products.

Some of these corrosion products will easily detach from the bundles when they are moved through the water. With fuel shuffle part 1 in progress, corrosion products will be deposited in the reactor cavity water as the bundles are removed from the core. Documentation of this is seen in the Operations Narrative Logs from PBAPS Refueling Outage 3R14. At 2032 on 9/21/2003, Fuel Shuffle Part 1 commenced. While this log entry does not specify this Fuel Shuffle as being Part 1, subsequent log entries at 2356 on 9/21/2003 and 0430 on 9/22/2003 confirm this as being Shuffle Part 1. The only activities scheduled to be performed in the reactor vessel during Shuffle Part 1 are fuel movements and some invessel visual inspection activities.

At 0122 on 9/23/2003, RHR Shutdown Cooling was removed from service temporarily for "fuel pool clarity". When RHR is in operation in normal Shutdown Cooling mode or Fuel Pool to Reactor Mode, the discharge of the operating RHR pump is to the bottom head area via the jet pump discharge.

The forced flow of water upward through the reactor tends to push corrosion products up, where RWCU has difficulty removing it. Removal of shutdown cooling is one option to allow the corrosion products to be drawn down to the RWCU pump suctions from the recirculation piping and the bottom head drain. This series of log entries shows a degradation in clarity as Fuel Shuffle Part 1 progresses. The effect is obviously worsened if RWCU trips and is out of service, since no removal of corrosion products will occur down in the reactor core area. The RWCU trip results in a reduction in filtration of the reactor cavity water, resulting in a higher concentration of corrosion products, and a degradation of the visibility of the reactor cavity water. Simply put, if the initial conditions assume a given amount of filtration, and some of that filtration is lost, less corrosion particles will be removed, and visibility will degrade. Furthermore, the only filtration system that may still be in service takes water only from the surface of the reactor cavity, spent fuel pool, and equipment pit. Any corrosion products in the water are required to travel to the surface to be removed. These particles diminish visibility due to the light from the installed spent fuel pool and reactor cavity lights reflecting off the particles, resulting in the appearance of a haze in the water. The effect worsens as the particle concentration increases. Since the fuel handling crew on the refuel platform is now required to look through more corrosion products in the water, visibility is degraded.

As shown on PBAPS P&ID M-363, sheet 1, water flows from the reactor cavity, spent fuel pool, and equipment pit to the skimmer surge tanks.

Since each of these three bodies of water have four returns to the skimmer surge tanks and are at the same height, an approximately equal amount of water flows from each area to the skimmer surge tanks. The typical fuel pool cooling alignment for Core Shuffle Part 1 is two or three fuel pool cooling pumps, heat exchangers, and demineralizers. Peach Bottom procedure SO 19.1 .A-2, Fuel Pool Cooling System Startup and Normal Operations, specifies in step 4.1.15.4 that a maximum flowrate of 550 gpm is permitted through each demineralizer. For example, if two demineralizers are in service, the maximum combined flowrate through the demineralizers is 1100 gpm. If SO 19.7.E-2, Aligning Fuel Pool Cooling System to Reactor Well, is performed, the return flow from the Fuel Pool Cooling System is split between the spent fuel pool and the reactor cavity.

Since both the spent fuel pool and reactor cavity each have two 6-inch returns, it can be assumed that approximately the same flow returns to each area.

This means that with two Fuel Pool Cooling pumps in service, 550 gpm would return to each area. This was the alignment for Fuel Shuffle Part 1 during the last PBAPS refueling outage, as shown in the attached Operations Narrative Logs from PBAPS Refueling Outage 2R15. The entry at 1 135 am on 9/16/2004 shows two Fuel Pool Cooling pumps, heat exchangers, and demins are in service, and aligned for return to both the spent fuel pool and reactor cavity per SO 19.7.E-2.

The RHR system alignment used during the entire Shuffle Part 1 for PBAPS 2R15 was Fuel Pool to Reactor Mode per A0 10.4-2, with a flowrate of 5000 gpm. This is a common mode, and depending on work that must be performed, can be the mode used for the majority of the outage. It is used extensively at both LGS and PBAPS. A0 10.4-2 aligns the operating RHR pump suction from the skimmer surge tanks, and discharges to the reactor vessel via the normal shutdown cooling discharge flowpath. The Narrative Logs show the RHR system was placed in this mode at 0401 am on 9/17/2004, approximately one hour before the start of Shuffle Part

1. RHR was maintained in this alignment until long after Shuffle Part 1 was completed.

Since the RHR pump is drawing 5000 gpm from the skimmer surge tanks, and the Fuel Pool Cooling system is drawing another 1100 gpm from the skimmer surge tanks, a total of 6100 gpm flows into and out of the skimmer surge tanks. Since about one-third of this flow (about 2000 gpm) is coming from the spent fuel pool, and only about 550 gpm of flow returning from the Fuel Pool Cooling system is returning to the spent fuel pool, then about 1450 gpm must flow from the reactor cavity to the spent fuel pool. The assumption that at least one-third of the water flowing into the skimmer surge tanks is from the spent fuel pool is a valid assumption, since the surface area of the spent fuel pool is slightly greater than one-third of the total surface area, and the weir plates will be adjusted to be consistent between the pools. According to Bill Bianco, Outage Services Engineer, surface areas of the three pools of water are as follows: Spent Fuel Pool - 616 sq. ft. Reactor Cavity - 550 sq. ft. Equipment Pit - 602 sq. ft. It is impossible for all 2000 gpm of the flow out of the spent fuel pool to come strictly from spent fuel pool water. Since only about 550 gpm is returning to the spent fuel pool from the Fuel Pool Cooling system, the only place the other 1450 gpm can come from is from the reactor cavity through the transfer canal. It is also not possible for more water to flow from the reactor cavity into the skimmer surge tanks than from the spent fuel pool. Per A0 10.4-2, step 4.1.4, the fuel pool to skimmer surge tank weir gates and reactor cavity to skimmer surge tank weir gates are in their lowest position. This is also required per SO 19.7.E-2. Since the weir plates in the reactor cavity match the level of the skimmer surge tank weirs on the spent fuel pool side, level would have to be higher in the reactor cavity to have higher flow. Since both bodies of water are connected through the transfer canal, it is not possible for them to be at different heights.

The significant amount of water flowing through the transfer canal from the reactor cavity into the spent fuel pool brings the degraded water from the reactor cavity into the spent fuel pool, causing its water to also degrade. Even when Fuel Pool Cooling is in service, the degradation will slowly worsen over time, as only about 370 gpm of the flow from the spent fuel pool (one-third of 1100) is filtered by the Fuel Pool Cooling demineralizers.

The attached Operations Narrative Logs from the most recent PBAPS Refueling outage (2R15) is provided in support of the above statements.

In summary, since the running RWCU pump yviJ trip on high motor winding temperature, and the CRD pump may or may not trip on overcurrent due to high temperature, reactor cavitykpent fuel pool level will either not change (if CRD trips), or will rise very slowly (if CRD does not trip).

In either case, this makes answer "d" incorrect. A loss of RWCU during Shuffle Part 1 will cause degradation of reactor cavity water, and with RHR in Fuel Pool to Reactor mode, which is a typical mode during refueling outages, spent fuel pool water visibility would also degrade. This makes answer "c" the correct answer. Answer "c" must be considered a valid answer, since if the exact same situation had actually occurred at any time during Shuffle Part 1 of the last PBAPS refueling outage, reactor cavity and spent fuel pool visibility would have degraded as a result of the RHR alignment being used, regardless of whether Fuel Pool Cooling was in service or not. Therefore, the Facility Licensee requests the correct answer for Question 37 be changed to "c".

~~ - References Provided: Design Basis Document P-S-09, Residual Heat Removal System Design Basis Document P-S-33, Reactor Building Closed Cooling Water System Design Basis Document P-S-36, Reactor Water Cleanup System Design Basis Document P-S-52, Fuel Pool Cooling and Cleanup System MCR ARC-211 , G-1 MCR ARC-215, A-2 and 6-2 PLOT-5003A, Control Rod Drive Hydraulic System Lesson Plan (PBAPS) PBAPS P&IDs M-361 sheet 1, M363 sheet 1 A0 10.4-2, Residual Heat Removal System - Fuel Pool to Reactor Mode SO 19.7.E-2, Aligning Fuel Pool Cooling System to Reactor Well SO 19.1 .A-2, Fuel Pool Cooling System Startup and Normal Operations Peach Bottom Archival Operations Narrative Logs for period of 9/20/2003 through 9/23/2003 Peach Bottom Archival Operations Narrative Logs for period of 9/15/2004 through 9/20/2004 To Skimmer Surge Tanks To Skimmer Surge Tanks To Skimmer Surge Tanks (-2000 gpm) A 1 Equipment Storage Pit 1 Reactor Cavity (Approx. 550 sq. ft.) ~~ (Approx. 602 sq. ft.) I450 gpm From From FPC RHR (550 gpm) (5000 gpm) Spent Fuel Pool (Approx. 616 sq. ft.) Skimmer Surge Tanks To RHR LGWPBAPS 2005 NRC LSRO Licensing Examination Question:

37 Page 37 of 50 PBAPS Unit 2 plant conditions are as follows: - Mode 5 - Core Shuffle Part I has just begun RBCCW is backing up TBCCW The CRD system is in service The RWCU system is in service in a normal lineup dumping 60 GPM to the Main A fire header break in the RBCCW room has caused both RBCCW pumps to trip - - - Condenser - WHICH ONE of the following describes the operational implications of this condition?

a. Higher than normal plant dose rates

b. Loss of Instrument Air to the Refueling Bridge c. Reactor Cavity and Fuel Pool visibility will degrade d. Reactor Cavity and Fuel Pool water level will begin to lower LGWPBAPS 2005 NRC LSRO Licensinq Examination Cognitive (H, L) Unit (0, 1, 2, 3) - -~_ I___~ 1 - - -----------i

-__--~--__ - __._____ _ Answer Key and Question Data - r - --- - - _____-- - ___ 1 Question # 37 i Choice I Basis or Justification

__-__ - 1 H PRA (Y/N) 1 LSRO 2 N IN Ja Source: Incorrect.

RWCU will not isolate on high temperature.

The isolation temperature is New Exam question 200°F but the Reactor Cavity temperature will be between 1 10°F and 130°F during refueling operations. Generally, the temperature is maintained well below 1 1 0°F. around 90°F. Incorrect.

The Refueling Bridge at PBAPS has an air compressor mounted on the bridge and is therefore independent of station air systems. Incorrect. RWCU will not isolate on high temperature.

The clarity of the Reactor Cavity will not change due to this event. Correct. With RBCCW supplying TBCCW loads, RBCCW is supplying cooling water to the CRD pump lube oil coolers and thrust bearings. The CRD pumps will trip after a loss of RBCCW. The loss of 60 GPM from the CRD system into the Reactor Cavity will cause Reactor Cavity and Fuel Pool levels to slowly lower. RWCU Dump flow is still in service and would lower cavity level at a rate of 60 gpm. This question tests differences between LGS and PBAPS Reference(s):

Learning None Required Attachments or Reference NLSRO-0370, M-316, GP-6, FH-6C NLSRO-0370 EO 1 L I I I 0 bjective:

KnowledgelAbility:

29501 8 AKI .01 j Importance:

3.6 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on componentkystem operations.

Prepared by: JBG PBAPS ALARM RESPONSE CARD WINDOW LOCATION ABCDEFGHJ ALARM WORDING B CRD WATER PUMP TRIP ~ ~___ ~___~___ - AUTOMATIC ACTIONS:

1. 2BP039, IIControl Rod Drive Water Pump Bff Trip. OPERATOR ACTIONS:

1. Verify pump trip at panel 20C005A. 2. Place pump control switch to flSTOP1l position.

3. Enter ON-107, ltLoss of CRD Regulating Function.Il CAUSE : 1. Low suction pressure.

2. Motor overcurrent.

ALARM SETPOINT:

PS-2-3-201B:

11" HG ABS (5.40 PSIA ABS) ACTUATING DEVICE ( S) : o PS-2-3-201B (Suction Pressure Switch) o 186-18 (E-42 Bus Differential Relay) o 127X-18 (E-42 Bus Undervoltage Relay) o 150/151 A, B, C (Motor Instantaneous/Timed o 150G (Ground Instantaneous Overcurrent Relay) o 186BX-18 (E-42 Bus Overcurrent Relay)

Overcurrent )

REFERENCES:

E-186 E-242 E-188 E-193 ALARM RESET AUTO ARC NUMBER: 211 20C205R G-1 ~ Rev. 2 PBAPS ALARM RESPONSE CARD WINDOW LOCATION ALARM WORDING ABCDE A CLEAN-UP RECIRC PUMP MTR WDG TEMP HIGH-HIGH AUTOMATIC ACTIONS:

2A RWCU Pump Trips. OPERATOR ACTIONS:

1. Verify Automatic Action.

2. Shutdown the RWCU System in accordance with SO 12.2.A-2, IIReactor Water Cleanup System Shutdown".

3. Investigate the cause of rising cooling water temperature/loss of cooling water. 4. IF the 2B RWCU Pump is available, THEN place RWCU in service with the 2B RWCU Pump in accordance with SO 12.1.A-2, IIReactor Water Cleanup System Startup for Normal Operations or Reactor Vessel Level Controlll . CAUSE : Decreased loss of RBCCW Cooling water to 2A RWCU Pump ALARM SETPOINT:

149 OF ALARM RESET: ACTUATING DEVICE (SI : AUTO TIS-2-12-089A

REFERENCES:

ARC NUMBER: 215 20C204R A-2 E-239 SO 12.2.A-2 E-368 SO 12.1.A-2 M-354 M-1-S-21 Rev. 3 PBAPS ALARM RESPONSE CARD I WINDOW LOCATION ALARM WORDING ABCDE B CLEAN-UP RECIRC PUMP MTR WDG TEMP HIGH - HIGH 2B RWCU Pump Trips. OPERATOR ACTIONS:

1. Verify Automatic Action. 2. Shutdown the RWCU System in accordance with SO 12.2.A-2, "Reactor Water Cleanup System Shutdown".

3. Investigate the cause of rising cooling water temperature/loss of cooling water.

4. IF the 2A RWCU Pump is available, THEN place RWCU in service with the 2A RWCU Pump in accordance with SO 12.1.A-2, "Reactor Water Cleanup System Startup for Normal Operations or Reactor Vessel Level Control".

CAUSE : Decreased loss of RBCCW Cooling water to 2B RWCU Pump ALARM SETPOINT:

149 OF ACTUATING DEVICE (S) : TIS-2-12-089B

REFERENCES:

E-239 SO 12.2.A-2 E-368 SO 12.1.A-2 M-354 M-1-S-21 ALARM RESET: AUTO ARC NUMBER: 215 20C204R B-2 Rev. 3 REACTOR WATER CLEANUP SYSTEM P-S-36 Revision 6 PECO Nuclear Peach Bottom Atomic Power Station, Units 2 and 3 Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD No. P-S-36 Page 1 of 120 1.2 SYSTEM DESCRIPTION The RWCU System (System Nos. 02, Reactor Recirculation System - (RPV bottom head drain line only), 12, Reactor Water Cleanup System, and 12A, RWCU Filter/Demineralizers) is a high pressure reactor water purification system for PBAPS Units 2 and 3. The RWCU System is classified as a primary power generation system. The RWCU System is designed to: - Maintain reactor water purity within specified limits by removing soluble and insoluble contaminants from the reactor coolant during the normal plant operating conditions of startup, power operation, hot standby, and shutdown (including refueling) 6 - Maintain reactor water level during plant startup, shutdown, and refueling by providing a blowdown path to discharge excess reactor water to the Main Condenser, Condensate Storage Tank (CST), or the Radwaste System (4.21) - Maintain circulation of reactor water when the Reactor Recirculation Pumps are unavailable to minimize temperature gradient and thermal stratification in the Reactor Recirculation piping and Reactor Pressure Vessel (RPV) - Automatically isolate upon receipt of Primary Containment Isolation System (XIS) isolation signals generated by Standby Liquid Control System (SLCS) initiation, low reactor water level, high RWCU System suction line flow, or high non- regenerative heat exchanger outlet temperature.

The RWCU System (System Nos. 02, 12) consists of two 100% capacity, motor-driven, vertical, sealless, centrifugal pumps arranged in parallel; one Regenerative Heat Exchanger (Regen HX) composed of three shell and tube heat exchangers connected in series; and two redundant Non-Regenerative Heat Exchangers (Non-Regen HX) each composed of two shell and tube heat exchangers connected in series. (4.32) (6 .l. 1.1) The RWCU Filter/Demineralizer (F/D) System (System No. 12A) is composed of two 50% capacity F/Ds along with a Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD No. P-S-36 Revision 6 Page 6 of 120 1.2 SYSTEM DESCRIPTION (continued) common regeneration subsystem which is able to backflush, precoat, and return a F/D to operation while the other F/D remains in service. The F/Ds purify the reactor water by mechanical filtration and ion exchange.

Periodic regeneration of a F/D is required due to depletion of the ion exchange resin and/or high F/D differential pressure.

The regeneration system consists of backflush connections for disposing of spent F/D resin and a common precoat tank and pump.

The backflush connections allow spent F/D resin to be discharged to the Radwaste System. Connections to the Low Pressure Air System allow low pressure air to be used as the motive force for F/D backflush operations.

The Condensate Storage and Transfer System supplies flush water to allow a thorough F/D backflush.

The precoat tank and pump are used to recoat the F/D internal filter elements with fresh resin slurry mixture after backflush operations.

Each F/D is provided with a holding pump which automatically starts upon sensing a low flow or whenever a F/D experiences a loss of flow condition. This ensures F/D resin is properly maintained on the filter elements.

A bypass line around the F/Ds is provided to control system flow while one or both F/Ds are out of service. (6.1.1.2) For normal system operation, one of the two 100% capacity RWCU Pumps take suction from the A Loop Reactor Recirculation System Pump suction line and from the RPV bottom head drain line through a common line penetrating Primary Containment.

The RWCU Pump discharge is cooled by passing it through the tube side I; f the Regen HX and then through the tube side of one of the two Non-Regen HXs. The cooled reactor water is directed to the two 50% capacity F/Ds for purification.

Outlet flow from the F/Ds is returned through the shell side of the Regen HX prior to being returned to the RPV via Reactor Core Isolation Cooling (RCIC) and Feedwater System piping.

(6.1.1.4)

During startup, shutdown, and refueling the outlet flow from the F/Ds can be discharged to the Main Condenser, CST, or Radwaste System in order to reduce excess RPV water level.

After regeneration of a F/D, F/D outlet flow is also discharged to the Main Condenser, CST or Radwaste in order ensure acceptable F/D effluent quality. The sealless RWCU Pumps are designed to handle radioactive reactor coolant at all normal reactor temperature and pressure operating conditions.

Each pump is encapsulated into a common pressure boundary Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD NO. P-S-36 Revision 6 Page 7 of 120 1.2 SYSTEM DESCRIPTION (continued) utilizing a common shaft and bearings, with an integral motor, pump support structure, and remote cooler. Thus the pump requires no shaft seals. Purge water, supplied by the CRD System, is provided to minimize the buildup of radioactive particles within the pump assembly.

The pump is provided with an internal thermal barrier to transfer heat from the hot reactor water for the pump motor cooler and the pump thermal coolant in the pump to the integral motor. Cooling barrier is provided by the Reactor Building Closed Cooling Water (RBCCW) System.

(6.1.1.9)

(6.1.2.1)

The Regen HX minimizes overall system heat losses by transferring the heat removed in the tube side flow from the RWCU Pumps to the shell side return flow from the F/Ds. The Non-Regen HX provides additional cooling of the Regen HX tube side outlet flow in order to protect the F/D ion exchange resins from excessive temperature.

Cooling water for the shell side of the Non-Regen HXs is provided by the RBCCW System. The standby RWCU Pump and redundant Non-Regen HX are provided to enhance RWCU System reliability and versatility. This enables the system to continue normal operation with one RWCU Pump or one Non-Regen HX out of service. The portions of the RWCU system which are classified as safety related are: - RWCU Pump suction line from the RPV and Recirculation System to the RWCU Pump Suction Primary Containment Isolation Valve (PCIV) MO- 2(3)-12-018 outside Primary Containment. - RWCU System return line from the RWCU Return PCIV M0-2(3)-12-068 to the RCIC System piping. - Two RWCU suction line differential pressure instrumentation lines penetrating Primary Containment out to and including their respective differential pressure indicator switches RWCU Break Isolation DP, DPIS-2(3)-12-124A and DPIS- 2 (3) 124B. - Two RWCU suction line flow instrumentation lines penetrating Primary Containment to their Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD No. P-S-36 Revision 6 Page 8 of 120 2.2 SYSTEM INTERFACES (continued) The Low Pressure Air System (System No. 36C) shall support operation of the RWCU System by providing low pressure air to backwash RWCU F/Ds 2(3)AF10 and 2(3)BF10 when needed for regeneration.

2.2.2.2.6 Instrument Air and Nitrogen Systems (6.1.13.6) The Instrument Air and Nitrogen Systems shall support operation of the RWCU System by providing clean, dry air from the Instrument Air System to the RWCU System air operated equipment to provide the force for valve operation.

2.2.2.2.7 Reactor Building Closed Cooling Water System (6.1.13.7)

The RBCW System shall support operation of the RWCU System by providing cooling water as required to the RWCU Pump motor coolers 2(3)AE455, 2(3)BE455, and the Non-Regen HXs during normal plant operation.

(4.32) 2.2.2.2.8 Reactor Core Isolation Cooling System (6.1.13.10)

The RCIC System shall support operation of the RWCU System by providing a RWCU flowpath to the Feedwater System to supply processed water to the RPV during startup, planned operation, and shutdown.

2.2.2.2.9 Radwaste System (6.1.13.9)

The Radwaste System shall support operation of the RWCU System by accepting contaminated or spent F/D resins and potentially contaminated liquids from the RWCU F/Ds and the RWCU F/D Precoat Tank during normal plant operation, and by accepting liquids collected from system vents and drains during all modes of plant operation.

2.2.2.2.10 Process Sampling System The Process Sampling System (System No. 12B) shall support operation of the RWCU by providing the capability for sampling and analyzing system liquids, during power operation and shutdown conditions, for purposes of making overall plant operational decisions.

The design permits in-line analysis or continuous Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD NO. P-S-36 Revision 6 Page 23 of 120 3.3 DESIGN FEATURES (continued)

BASIS: High differential pressure corresponds to high flow in the RWCU Pump suction line which is indicative of pipe rupture in the RWCU piping.

The high differential pressure isolation of the RWCU Systems prevent reactor water inventory loss to meet the design inputs of AEC Criterion 12 (2.4.1.2.1.7), AEC Criterion 51 (2.4.1.2.1.11}, and System Interface (2.2.1.1.1).

3.3.1.4.3 RWCU F/D High Differential Pressure Isolation A high differential pressure across the RWCU F/D or its discharge strainer automatically isolates the respective F/D by closing the corresponding RWCU F/D outlet valve. RWCU F/D Differential Pressure Transmitter DPT-2(3)-12-4-069A detects differential pressure across the RWCU F/D and transmits a signal to RWCU F/D Differential Pressure Switch DPS-2(3)-12 082A. RWCU F/D Post Strainer Differential Pressure Switch DPIS-2(3)-12-4-072A detects differential pressure across the RWCU Post Strainer 2(3)AF065.

These differential pressure switches send a signal to close valve CV-2-12A-016A when the differential pressure of either switch exceeds the setpoint.

The high differential pressure is an indication of a clogged filter or strainer. Loop B is similar to loop A. (6.1.1.2)

(6.1.1.24, Sh 2) BASIS: High differential pressure isolates the corresponding F/D to protect the F/D from damage due to high flow to prevent fouling of the F/D elements, and to prevent resin material carry over into the Reactor to meet the design inputs of AEC Criterion 12 {2.4.1.2.1.7}, and System Protection (2.5.21. 3.3.1.4.4 RWCU Pump Trips Any one or combination of conditions listed below trip the RWCU Pumps 2(3)A049 and 2(3)B049. - RWCU Inlet Inboard PCIV MO-2(3)-12-015 not fully - RWCU Inlet Outboard PCIV M0-2(3)-12-018 not fully - RWCU Pump Motor Winding Temperature High-High - RWCU Pump Motor overload.

(4.24) (4.32) (6.1.1.1) (6.1.1.22) open open Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD NO. P-S-36 Revision 6 Page 67 of 120 3.3 DESIGN FEATURES (continued)

BASIS: The RWCU Pumps trip on closure of the RWCU PCIVs to protect the pump from damage due to continued operation with no suction source to meet the design inputs of System Protection (2.5.2). High RWCU Pump Motor Winding temperature is an indication of insufficient cooling for the RWCU Pumps. The high RWCU Pump Motor winding temperature and pump motor overload trips protect the RWCU Pumps from damage due to overheating and excessive loading during off- normal transient events to meet the design inputs of System Protection (2.5.2). 3.3.1.4.5 RWCU F/D Outlet Flow Trip/Isolation RWCU F/D Holding Pumps 2 (3)AP053 and 2 (3)BP053 trip on power failure, when the flow through the RWCU F/D returns to normal. (6.1.1.2)

(6.1.1.24, Sh 2) BASIS: The RWCU F/D normal flow is sufficient to prevent dislodging of the resin coating on the filters and operation of holding pump is not required.

Tripping the holding pump on RWCU F/D normal flow protects the holding pump from unnecessary operation to meet the design inputs of System Protection (2.5.2).

3.3.1.4.6 RWCU F/D Precoat Pump Shutoff RWCU F/D Precoat Tank Level Switch LS-2(3)-12-4-080 provides signal to shutoff or prevent startup of RWCU Precoat Pump 2(3)0P051 when the RWCU F/D Precoat Tank 2(3)0T020 level is below the low level setpoint.

The RWCU F/D Precoat Tank Level Switch LS-2(3)-12-4-080 also provides a signal to shutoff or prevent startup of the RWCU Precoat Pump 2(3)0P051 when the RWCU F/D Precoat Tank 2(3)0T020 is above the high level setpoint.

This is a result of a "HALT" function from XIC-2(3)-12-4-097 which occurs during F/D regeneration.

(6.1.1.2)

(6.1.1.24, Sh 2) BASIS: The RWCU Precoat Pump trips on low level to protect the pump from damage due to continued operation with low suction head to meet the design inputs of System Protection (2.5.2). Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Water Cleanup System DBD No. P-S-36 Revision 6 Page 68 of 120 FUEL POOL COOLING AND CLEANUP SYSTEM P-S-52 Revision 5 PECO Nuclear Peach Bottom Atomic Power Station, Units 2 and 3 Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD NO. P-S-52 Page 1 of 79 FUEL POOL COOLING AND CLEANIlP SYSTEM P-S-52 REVISION CONTROL 1 2 Rev. I I I I I on 12/5/96 This Issue Incorporates ECR #'s 95-04378, 95-05195 and 95-02328, 95-03313, 95-05196 6/4/98 This Issue Incorporates ECR #;s 95-05197 R1, 95-05450 R1, 96-03575 R1, 98- No. I Date !Reason for Issue I Prepared I Reviewed I Approved 01 6/23/95 I Original Issue I See archived copies for signatures 00712 RO This issue incorporates ECRs 97-02488, Rev. 1 and 97- 002934-00 Rev.

1 This issue incorporates ECR 99- 00025, Rev.

3. This issue incorpor a- os bCP# s 01-01188 Re\'. 0, 01- 01200 Re-"?. 0, 02-00016 Rev. 0, 02-00314 Rev. 0 & 03-00443 Pev. 0 DM? Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System last revisions.

DBD NO. P-S-52 Page 2 of 79 FUEL POOL COOLING AND CLEANUP SYSTEM DESIGN BASELINE . TABLE OF CONTmS SECTION 1.0 1.1 1.2 1.3 2.0 2.1 2.2 2.3 2.4 2.5 3.0 3.1 3.2 3.3 4.0 5.0 6.0 6.1 6.2 6.3 FIGURES I.RO..ION

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4 SCOPE AND LIMITATIONS

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4 SYSTEM DESCRIPTION

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6 DEFINITIONS

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8 DESIGN INPUTS

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10 SYSTEM ..IES ....................................

10 SYSTEM I.E.ACES ....................................

18 EXTERNAL INFLUENCES ON SYSTEM DESIGN .................

22 REQUIRemENTS. COMMITMENTS.

CODES AND STANDARDS

....... 25 OTHER DESIGN INPUTS

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32 SYSTEM DESIGN BASELINE ...............................

34 SYSTEM F[TNCTIONS

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34 CONTROLLING PANMETERS

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37 DESIGN FEATURES ......................................

41 DESIGN WELINE EVOLUTION

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66 DIFFE-ES BETWEEN UNITS

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70 REFEREhFCES

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71 CONTROLLED .S .................................

71 REFERENCE BOOK (UNCONTROLLED DOWNENTS)

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78 SYSTEM I.. .........................................

79 TABLES T2.1-1-DBD BOUNDARIES ELECTRICAL POWER NONE USED 2 PAGES APPENDICES NONE USED

1.0 INTRODUCTION

1.1 SCOPE AND LIMITATIONS This design baseline document (DBD) provides the design bases for the Fuel Pool Cooling and Cleanup System at Peach Bottom Atomic Power Station, Units 2 and 3. System functions, controlling parameters, and design features for normal, abnormal, and accident conditions including precautions and limitations are addressed in this document as they relate to the Fuel Pool Cooling and Cleanup System.

In addition, regulatory requirements, commitments, codes and standards, and system configuration changes that had impact on the design baseline of the system have been identified and discussed.

This DBD for the Fuel Pool Cooling and Cleanup System has been developed to assist in future design and evaluation activities, licensing activities, 10CFR50.59 reviews, and as the primary reference for preparing and updating training material. In addition, the DBD, when used in conjunction with applicable other documents (e.g., Technical Specifications, UFSAR, procedures, etc.) may assist in making operability determinations.

This DBD is a comprehensive system level document that describes regulatory requirements applicable to the Fuel Pool Cooling and Cleanup System as well as other requirements for design of the Fuel Pool Cooling and Cleanup System. This DBD is not, however, considered the only document that should be reviewed in performing design and evaluation activities that impact the Fuel Pool Cooling and Cleanup System.

The contents of this DBD focus on the requirements of this system as bounded by the DBD boundaries.

Items that are considered good practices are not necessarily included. Detailed design information which exists in other controlled documents (specifically design output documents and procedures) is not necessarily repeated in this DBD and will have to be used in conjunction with this document. This DBD is presented in six major sections.

Section 1 provides an introduction to and description of the basic functions of the system. The information overviews the design basis of the system. Section 2 identifies all design inputs to the system including the regulatory, commitment, code and standard inputs, as well as inputs from interfacing systems and/or topical areas. The system boundaries for the DBD discussion are also identified in this section. The information in section 2 is considered design 1.1 SCOPE AND LIMITATIONS (continued) input, both required and self-imposed, to the Fuel Pool Cooling and Cleanup System. Section 3 describes the system design baseline. This section includes discussions on system functions, safety classifications, controlling parameters and associated design margins, and design features including controls philosophy, instrumentation loops, and subsystems.

Design baseline information includes an internal reference to a design input source identified in section 2. Design detail that has been included typically results from such things as vendor imposed requirements, the design of the system, etc. Section 4 summarizes the changes to the DBD content which have occurred since the plant was licensed. This information by itself is not considered design basis information.

The design basis information related to the impact of plant changes on system design basis is discussed in sections 2 and 3 of the DBD. Section 5 is a summary listing of system differences between units. This information by itself is not considered design basis information.

Design basis information related to system differences is discussed in sections 2 and 3 of the DBD. Section 6 is a listing of Reference Documents. This information is not considered design basis information. The DBD provides references to the source documents that specifically support the design basis information contained herein. It should be noted that this DBD does not recreate the step-by-step detailed design process or otherwise describe the iterative process used in the detailed plant design phase.

The DBD does not provide the answer to questions regarding the function and design history of the system hardware.

Therefore, the user should not assume that this DBD is the single source of all information for Fuel Pool Cooling and Cleanup System.

References to equipment, both by equipment ID and description, utilized in this DBD are based on equipment IDS and descriptions contained within the Plant Information Management System (PIMS). Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD NO. P-S-52 Revision 5 Page 5 of 79 1.2 SYSTEM DESCRIPTION The Fuel Pool Cooling and Cleanup System (System Nos. 19 and 19A) is a cooling and cleanup system for Peach Bottom Atomic Power Station, Units 2 and 3. For the purposes of this DBD, the Fuel Pool Cooling and Cleanup System consists of the following systems: - Fuel Pool Cooling (System No. 19) - Fuel Pool Filter Demineralizer (System No. 19A). These system descriptions are provided below:

The non-safety related Fuel Pool Cooling and Cleanup System is designed to remove the decay heat generated by the spent fuel assemblies stored in the fuel pool and to maintain the pool water at a clarity and purity suitable both for underwater operations and for the protection of personnel in the refueling area. Each fuel pool is provided with a Fuel Pool Cooling and Cleanup System.

In addition, a spare Filter Demineralizer (F/D) is common to both units fuel pool. The Fuel Pool Cooling (System No. 19) consists of the following major components: - 3 Fuel Pool Cooling Pumps (2 (3)AP041, 2 (3)BP041 and 2 (3)CP041) - 3 Fuel Pool Heat Exchangers (2 (3IAEO20, 2 (3IBE020 and 2 (3 ) C-EO20) - 2 Fuel Pool Skimmer Surge Tanks (2(3)AT016 and 2(3)BT016).

The Fuel Pool Filter Demineralizer (System No. 19A) consists of the following major components: - 3 Fuel Pool Filter Demineralizers (OAF008, OBF008 and OCF008) - 3 Fuel Pool Filter and Demin Holding Pumps (OAP086, OBP086 and OCP086) - Waste Precoat Tank (00T056) - Fuel Pool/Radwaste Precoat Pump (OOP032).

The Fuel Pool Cooling and Cleanup System removes decay heat from fuel stored in the Spent Fuel Pool and 1.2 SYSTEM DESCRIPTION (continued) includes equipment to maintain the purity of the water in the system. Water from the Spent Fuel Pool flows through weirs and a wave suppression scupper at the pool surface into two skimmer surge tanks adjacent to the pool. Water in the skimmer surge tanks flows by gravity through the fuel pool heat exchangers to the suction of the fuel pool cooling pumps. From the pumps, water is returned to the Spent Fuel Pool through two discharge lines located near the top of the fuel racks. The discharge flow of the pumps is diverted through the cleanup loop before being returned to the pool. Three centrifugal pumps and heat exchangers are provided for circulating and transferring heat from the fuel pool water to the Service Water System. The number of pumps and heat exchangers operated are dependent on the heat load.

The Filter Demineralizer in the cleanup loop maintains pool water purity and clarity by a combination of filtration and ion exchange. Disposable ion exchange resins in the filter demineralizer remove ionic fission product and corrosion product impurities and also serve as a filter for particulate matter.

The cleanup loop includes a Filter Demineralizer for each unit located separately in shielded cells in the Radwaste Building and a spare Filter Demineralizer common to the two reactor units. The Fuel Pool Filter Demineralizer is a precoat-type, using powdered cation-anion resins as the coating media on the external surface of the filter elements.

The filter elements are cylindrical stainless steel mesh, mounted vertically in a tube sheet and replaceable as a unit. The ion exchange resin is a mixture of finely ground cation and anion resins. This resin is referred to as precoat.

The precoat is applied to the surface of the filter elements by a flowing process called precoating.

A strainer is provided in the effluent stream of each Filter Demineralizer to protect against catastrophic failure of a filter element. The backwash and precoat subsystem is common to the two reactor units and serves all three Filter Demineralizers. Included in the subsystem are a precoat tank and filter precoat pump. New ion exchange resin is mixed in the precoat tank and transferred as a slurry by the filter-precoat pump to the Filter Demineralizer, where it is deposited on the filter elements.

An agitator is provided with the precoat Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD No. P-S-52 Revision 5 Page 7 of 79 1.2 SYSTEM DESCRIPTION (continued) tank for mixing. The precoat subsystem can also be used for cleaning the Filter Demineralizers.

During normal plant operation, the Fuel Pool Cooling and Cleanup System serves only the Spent Fuel Pool. During refueling operations, however, when the Reactor Well and Dryer/Separator Storage Pit, are filled with water, the Fuel Pool Cooling and Cleanup System can be aligned to recirculate and process the water in all these cavities. Water from the Refueling Water Storage Tank or the Condensate Storage Tank is used to fill the refueling area cavities.

The refueling water pumps fill the Reactor Well and the Dryer/Separator Storage Pit through diffusers in the Reactor Well. After refueling activities are completed, the refueling water pumps transfer water from the refueling area cavities back to the Refueling Water Storage Tank via a condensate filter demineralizer if additional cleanup is required. Gravity draining of the refueling water directly to the Refueling Water Storage Tank is also possible.

As the heat load on the Spent Fuel Pool changes, the number of operating fuel pool cooling pumps and heat exchangers is adjusted to maintain the desired water temperature.

The Fuel Pool Cooling and Cleanup System has sufficient cooling capacity to maintain the Spent Fuel Pool water at a temperature at or below 150F, for a normal decay heat load with two pumps and two heat exchangers operating.

If an abnormally large heat load is placed in the Spent Fuel Pool, a cooling train of the RHR System, consisting of an RHR pump and heat exchanger, is substituted for the Fuel Pool Cooling pumps and heat exchangers for cooling the pool water. The conditions under which cooling of the Spent Fuel Pool water by the RHR System alone would be required include the unloading of a full core load of irradiated fuel into the pool. Alignment of the RHR System to the Fuel Pool Cooling System requires manual operator action. If the normal systems used for Spent Fuel Pool makeup are unavailable, fire hoses can be used as a source of makeup water. (4.4) (6.1.1.1)

(6.1.1.2)

(6.1.6.1)

(6.1.7.4) 1.3 DEFINITIONS Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD NO. P-S-52 Revision 5 Page 8 of 79 1.3 DEFINITIONS (continued) Definitions provide the DBD user a common reference for understanding terms used within the DBD. Definitions, if provided here, shall be used in conjunction with the definitions contained in CNG AA-CG-2. Additionally, procedure NE-C-230-8 provides definitions which apply to all DBDs.

1.3.1 None Used. Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD NO. P-S-52 Revision 5 Page 9 of 79 2.2 SYSTEM INTERFACES (continued)

The Fuel Pool Cooling and Cleanup System requires no support from other plant systems to support Technical Specification operability.

2.2.2.2 Other Supporting Systems Generally, the Fuel Pool Cooling and Cleanup System Technical Specification operability is not supported by operation of the following systems.

However, each referenced DBD reviewed in conjunction with other applicable documents (e.g., UFSAR, Technical Specifications, procedures, etc.) may assist in determining operability for each situation requiring a determination.

2.2.2.2.1 Residual Heat Removal System (6.1.13.2)

The Residual Heat Removal System shall support operation of the Fuel Pool Cooling and Cleanup System by providing supplemental heat removal capability for cooling the fuel pool when needed during refueling.

The Fuel Pool Cooling and Cleanup System is designed to meet the cooling requirements for most fuel pool heat loads and system configurations.

However, additional heat removal capability may be needed when full core off-loading occurs and less than three Fuel Pool Cooling pumps/heat exchangers are available.

The Residual Heat Removal System shall also support operation of the Fuel Pool Cooling and Cleanup System by recording fuel pool temperature during all reactor operating modes whenever fuel is in the fuel pool. 2.2.2.2.2 Service Water System (6.1.13.4)

The Service Water System shall support operation of the Fuel Pool Cooling and Cleanup System by providing cooling water at a flow rate of 800 GPM and a maximum temperature of 9OF to each of the Fuel Pool Cooling Heat Exchangers during normal plant operation when offsite power is available.

2.2.2.2.3 Instrument Air and Nitrogen Systems (6.1.13.5)

The Instrument Air and Nitrogen System shall support operation of the Fuel Pool Cooling and Cleanup System Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD NO. P-S-52 Revision 5 Page 20 of 79 2.4 REQUIREMENTS, COMMITMENTS, CODES AND STANDARDS (continued) 2.4.1.2.1.3 AEC Criterion 67, Fuel and Waste Storage Decay Heat (Category B) (6.1.11.4) "Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs." The Fuel Pool Cooling and Cleanup System shall be designed to provide reliable decay heat removal to the Spent Fuel Pool to conform with AEC Criterion 67 as documented in UFSAR, Appendix H (6.1.7.1).

(3.1.1) (3.3.2.1.7) 2.4.1.2.1.4 AEC Criterion 68, Fuel and Waste Storage Radiation Shielding (Category B) (6.1.11.4) "Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10CFR20." The Fuel Pool Cooling and Cleanup System shall be designed to maintain adequate radiation shielding to conform with AEC Criterion 68 as documented in UFSAR, Appendix H (6.1.7.1). (3.3.2.1.2) (3.3.2.1.3) 2.4.1.2.2 Updated Final Safety Analysis Report, Section 10.5, Fuel Pool Cooling and Cleanup System (6.1.7.4)

This UFSAR Section provides the following criteria: - To minimize corrosion product buildup and control water clarity through filtration and demineralization - To minimize fission product concentrations which could be released from the pool water to the reactor building environment - To monitor fuel pool water level and maintain a water level above the fuel sufficient to provide shielding for normal building occupancy Peach Bottom Atomic Power Station, Units 2 and 3 Fuel Pool Cooling and Cleanup System DBD NO. P-S-52 Revision 5 Page 30 of 79 3.0 SYSTEM DESIGN BASELINE Section 3.0 provides the system design baseline.

The system design baseline identifies how the system fulfills the design inputs identified in section

2. For every specific design input in section 2.2.1, 2.3.2, 2.4, and 2.5 there is a corresponding discussion in section 3. All discussions in section 3 are traceable to the associated section 2 design inputs via internal DBD referencing. Section 3.1 identifies the system functions and the associated alignments for these functions.

The bases for the system functions are provided via internal "in the text" references to section 2 design inputs.

Section 3.2 identifies the controlling parameters (numerical values) associated with each of the system functions.

This includes controlling parameters associated with subsystems and major pieces of equipment. For each controlling parameter, the controlling source document (with external DBD references) and the associated design margin are identified.

Section 3.3 identifies the remaining design features associated with the system. This includes design features associated with subsystems and major pieces of equipment.

The bases for these design features are provided via internal "in the text" references to section 2 design inputs. 3.1 SYSTEM FUNCTIONS Section 3.1 identifies two system functions for the Fuel Pool Cooling and Cleanup System: Spent Fuel Decay Heat Removal and Maintain Fuel Pool Water Quality and Clarity. The alignments, including alternatives, are identified for this system function.

The Fuel Pool Cooling and Cleanup System primary function is to remove decay heat from fuel stored in the Spent Fuel Pool. The Fuel Pool Cooling and Cleanup System performs the decay heat removal function whenever spent fuel is stored in the Spent Fuel Pool, including refueling. During refueling operations when the Reactor Well and Dryer/Separator Storage Pit, are filled with water, the Fuel Pool Cooling and Cleanup System can be aligned to recirculate and process the water in these cavities.

REACTOR BUILDING CLOSED COOLING WATER SYSTEM P-s-33 Revision 8 PECO Nuclear Peach Bottom Atomic Power Station, Units 2 and 3 Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD No. P-S-33 Page 1 of 69 REACTOR BUILDING CLOSED COOLING WATER SYSTEM P-s-33 REVISION CONTROL Reason for Issue Initial Issue IncorDorates ECR # 93- Rev. No. 0 1 2 3 4 5 6 7 Prepared Reviewed Approved 8 This issue incorporates This issue incorporates ECR 98-03212, Rev. 0 ECR 98-01007 Rev 0 Date 7/26/93 10/7/93 BC B DEP JAJ 8/4/94 12 / 6 / 94 6/13/95 9/11/95 6/8/98 9/22/99 12/6/99 - 2021 Issue incorporates ECR #'s 93-03828; 94-05152; 94-06449: 94-06751 and ~~ ~ 94-06993' I I I Issue incorDorates ECR I #'s 94-08123; 94-08233; 94-08823; 94-09052 and Issue incorporates ECR 94-10231; 94-10458 and This issue incorporates 94-09619 #'s 93-03802; 94-08660; 94-11752 ECR #'s 94-07188; 95-1695; 95-02070; 95- *02271 and 95-03022 This issue incorworates ECR #'s 94-05157-Rev 0; 96-03635 Rev 0; 96-04016 Rev. 0; and 98-00712 Rev n I I Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD No. P-S-33 Page 2 of 69 SECTION 1.0 1.1 1.2 1.3 2.0 2.1 2.2 2.3 2.4 2.5 3.0 3.1 3.2 3.3 4.0 5.0 6.0 6.1 6.2 6.3 FIGURES REACTOR BUILDING CLOSED COOLING WATER SYSTEM DESIGN BASELINE . TABLE OF CO-S PAGE INTRODUCTION

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4 SCOPE AND LIMITATIONS

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4 SYSTEM DESCRIPTION

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6 DEFINITIONS

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9 DESIGN I..S ........................................

10 SYSTEM BOUNDARIES

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10 SYSTEM INTERFACES

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17 EXTERNAL INFLUENCES ON SYSTEM DESIGN .................

22 REQUI-S. COMMITMENTS.

CODES AND STANDARDS

....... 25 OTHER DESIGN IrJPUTS ..................................

34 SYSTEM DESIGN BASELINE ...............................

36 SYSTEM FUNCTIONS

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36 CONTROLLING PARAMETERS

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40 DESIGN FEATURES ......................................

46 DESIGN BASELINE EVOLUTION

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59 DIFFERENCES BETWEEN UNITS

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62 REFERENCES

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63 CONTROLLED DOCUMENTS

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63 REFERENCE BOOK (UNCONTROLLED -SI ..............

69 SYSTEM INDEX .........................................

70 TABLES T2.1-1 DBD BOUNDARIES . ELECTRICAL POWER APPENDICES Peach Bottom Atomic Power Station.

Units 2 and 3 Reactor Building Closed Cooling Water System NONE USED 2 PAGES NONE USED DBD No . P-S-33 Revision 8 Page 3 of 69 1.0

1.1 INTRODUCTION

SCOPE AND LIMITATIONS This design baseline document (DBD) provides the design bases for the Reactor Building Closed Cooling Water System at Peach Bottom Atomic Power Station, Units 2 and 3. System functions, controlling parameters, and design features for normal, abnormal, and accident conditions including precautions and limitations are addressed in this document as they relate to the Reactor Building Closed Cooling Water System.

In addition, regulatory requirements, commitments, codes and standards, and system configuration changes that had impact on the design baseline of the system have been identified and discussed.

This DBD for the Reactor Building Closed Cooling Water System has been developed to assist in future design and evaluation activities, licensing activities, 10CFR50.59 reviews, and as the primary reference for preparing and updating training material.

In addition, the DBD, when used in conjunction with applicable other documents (e.g., Technical Specifications, UFSAR, procedures, etc.) may assist in making operability determinations. This DBD is a comprehensive system level document that describes regulatory requirements applicable to the Reactor Building Closed Cooling Water System as well as other requirements for design of the Reactor Building Closed Cooling Water System. This DBD is not, however, considered the only document that should be reviewed in performing design and evaluation activities that impact the Reactor Building Closed Cooling Water System.

The contents of this DBD focus on the requirements of this system as bounded by the DBD boundaries. Items that are considered good practices are not necessarily included. Detailed design information which exists in other controlled documents (specifically design output documents and procedures) is not necessarily repeated in this DBD and will have to be used in conjunction with this document. This DBD is presented in six major sections.

Section 1 provides an introduction to and description of the basic functions of the system.

The information overviews the design basis of the system. Section 2 identifies all design inputs to the system including the regulatory, commitment, code and standard inputs, as well as inputs from interfacing systems Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD NO. P-S-33 Revision 8 Page 4 of 69 1.1 SCOPE AND LIMITATIONS (continued) and/or topical areas. The system boundaries for the DBD discussion are also identified in this section. The information in section 2 is considered design input, both required and self-imposed, to the Reactor Building Closed Cooling Water System. Section 3 describes the system design baseline.

This section includes discussions on system functions, safety classifications, controlling parameters and associated design margins, and design features including controls philosophy, instrumentation loops, and subsystems. Design baseline information includes an internal reference to a design input source identified in section

2. Design detail that has been included typically results from such things as vendor imposed requirements, the design of the system, etc. Section 4 summarizes the changes to the DBD content which have occurred since the plant was licensed.

This information by itself is not considered design basis information.

The design basis information related to the impact of plant changes on system design basis is discussed in sections 2 and 3 of the DBD. Section 5 is a summary listing of system differences between units. This information by itself is not considered design basis information. Design basis information related to system differences is discussed in sections 2 and 3 of the DBD. Section 6 is a listing of Reference Documents. This information is not considered design basis information.

The DBD provides references to the source documents that specifically support the design basis information contained herein. It should be noted that this DBD does not recreate the step-by-step detailed design process or otherwise describe the iterative process used in the detailed plant design phase.

The DBD does not provide the answer to questions regarding the function and design history of the system hardware.

Therefore, the user should not assume that this DBD is the single source of all information for Reactor Building Closed Cooling Water System. References to equipment, both by equipment ID and description, utilized in this DBD are based on equipment IDS and descriptions contained within the Plant Information Management System (PIMS). Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD No. P-S-33 Revision 8 Page 5 of 69 1.2 SYSTEM DESCRIPTION 4 The Reactor Building Closed Cooling Water (RBCCW) System (System No. 35) is an non-safety related auxiliary system for the Peach Bottom Atomic Power Station, Units 2 and 3. The Reactor Building Closed Cooling Water System is designed to perform the following functions: - To provide cooling water to remove the maximum anticipated heat loads developed by the components served by the system over the full range of normal plant operating conditions and ambient temperature conditions - To operate during normal plant operation and on a LOSS of Offsite Power (LOOP) - To serve as a barrier between potentially radioactive systems and the Service Water System. The RBCCW System consists of two 100% capacity cooling water pumps, 2(3)AP010 and 2(3)BP010, two 100% capacity heat exchangers, 2 (3)AE018 and 2 (3)BE018, one head tank, 2(3)0T201, one chemical addition tank, 2(3)0T202, and associated valves, piping, and controls. During normal plant operation, one RBCCW Pump and one RBCCW Heat Exchanger are in service.

The second pump automatically starts on low pressure in the supply header, supplying additional flow through the heat exchanger in operation. During normal plant operation, the RBCCW System provides cooling water to the following components: Reactor Water Cleanup (RWCU) Non-Regenerative Heat Exchangers RWCU Recirculation Pump Seal Coolers Reactor Recirculation Pump Seal and Motor Oil Coolers Post Accident Sampling System Coolers Sample Station Coolers Reactor Building Equipment Drain Sump Cooler Waste Filter Holding Pump Cooler Floor Drain Filter Holding Pump Cooler Material Test Stations Instrument Nitrogen Compressors and Aftercoolers.

Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD No. P-S-33 Revision 8 Page 6 of 69 1.2 SYSTEM DESCRIPTION (continued)

The cooling water is circulated throughout the closed- loop system by the RBCCW Pumps. The heat gained from the components being cooled is transferred to the Service Water System through the RBCCW Heat Exchangers.

The RBCCW Pump Motors are connected to Class 1E busses. The RBCCW System also has the capability to supply cooling water to the Fuel Pool Heat Exchangers in the event that the Service Water System is not available.

The supply of this water is through spectacle flanges and spool pieces which are normally removed. In the event of a LOOP, the RBCCW Pump which was running automatically restarts when power is restored to its Class 1E bus. The RBCCW Pump which was in "AUTO" will automatically start after a predetermined time if RBCCW discharge header pressure is not reestablished by the pump which was running. During a LOOP, the RBCCW System supply to the following components is isolated: - RWCU Non-Regenerative Heat Exchangers - RWCU Recirculation Pump Seal Coolers - Sample Station Coolers 2 (3) OS107 and 2 (3) OS113 - Material Test Stations - Instrument Nitrogen Compressors and Aftercoolers.

The RBCCW System will then provide cooling water to the following components normally supplied by RBCCW: - Reactor Recirculation Pump Seal and Motor Oil - Post Accident Sampling System Coolers - Sample Station Cooler OOS106 - Reactor Building Equipment Drain Sump Cooler - Waste Filter Holding Pump Cooler - Floor Drain Filter Holding Pump Cooler. Coolers In addition, the Drywell Area Cooling Coils, the Drywell Equipment Sump Cooler, and the Recirculation Pump Motor Air Coolers, normally served by the Chilled Water System, and the CRD Pump Oil Coolers and the Service and Instrument Air Compressors, normally served by the Turbine Building Closed Cooling Water (TBCCW) System, are supplied with cooling water from the RBCCW System. Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD NO. P-S-33 Revision 8 Page I of 69 1.2 SYSTEM DESCRIPTION (continued)

In the event of a loss of power to two of the three Drywell Chillers for a predetermined period of time, the RBCCW supply to various components will be isolated in the same manner as occurs during a LOOP. The RBCCW System supply of cooling water to the Drywell Area Cooling Coils, the Drywell Equipment Sump Cooler, and the Recirculation Pump Motor Air Coolers would then be utilized.

In the event of a loss of power to both TBCCW Pumps for a predetermined period of time or if both TBCCW Pumps are stopped for a predetermined period of time, the RBCCW interconnect valves to the TBCCW System would be opened and the RBCCW System would supply cooling water to the CRD Pump Oil Coolers and to the Service and Instrument Air Compressors.

I A piping interconnection with the ESW System exists which would allow ESW System cooling water to be supplied to the RBCCW Heat Exchangers.

The interconnecting valves are locked closed because the RBCCW System has not been seismically qualified to be connected to a safety related system and due to the adverse hydraulic effects to safety related components served by ESW. Therefore, no heat sink is available to the RBCCW System in the event of a LOOP or loss of the Service Water supply.

Makeup water to the RBCCW System is supplied to the RBCCW Head Tank by the Makeup and Demineralized Water System. The tank provides a constant head to maintain RBCCW Pump NPSHA and an accumulator to respond to temperature changes in the system. Chemicals can be added to the RBCCW System through the RBCCW Chemical Addition Tank for corrosion prevention throughout the system. A radiation monitor is provided in the RBCCW recirculation line to indicate, record, and alarm the presence for radioactivity in the RBCCW System.

(6.1.1.1)

(6.1.1.2)

(6.1.7.3)

Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD No. P-S-33 Revision 8 Page 8 of 69 2.2 SYSTEM INTERFACES (continued)

Reactor Building Material Test Station during normal plant operation and during a LOOP. (3.1.1) (3.3.1.1.1)

(3.3.1.1.2)

(3.3.1.2.1) (3.3.1.3.1) (3.3.1.5.1)

(3.3.2.1.1)

(3.3.2.1.2)

(3.3.4.5)

(5.1) (5.2) 2.2.1.2.5 Post Accident Sampling System (6.1.13.7)

The RBCCW System shall support operation of the PASS by providing cooling water as required to Unit 2 and Unit 3 PASS Sample Coolers E-604 and E-605 during normal plant operation and during a LOOP. (3.1.1) (3.3.1.1.1)

(3.3.1.1.2)

(3.3.1.2.1)

{3.3.1.3.1)

(3.3.1.3.2)

(3.3.1.5.1) (3.3.2.1.1) (3.3.2.1.2)

(3.3.4.5) 2.2.1.2.6 Turbine Building Closed Cooling Water System (6.1.13.25)

The RBCCW System shall support operation of the TBCCW System by providing up to 72 GPM of cooling water at a maximum of 1OOF to the operating CRD Pump Lube Oil Cooler and Thrust Bearing Housing, and the two (out of four) operating air compressors low pressure and high pressure water jackets, intercoolers, aftercoolers, oil coolers, and bleed off coolers only. This support shall be available during a LOOP or whenever both TBCCW Pumps are unavailable for service. (3.1.1) (3.3.1.1.1) (3.3.1.1.2) (3.3.1.2.1)

(3.3.1.3.1)

(3.3.1.3.2)

(3.3.1.5.1) (3.3.2.1.1) (3.3.2.1.2)

(3.3.4.5) 2.2.1.2.7 Radwaste System (6.1.13.9)

The RBCCW System shall support operation of the Radwaste System by providing cooling water as required to the Waste Filter Holding Pump Cooler OOE108, the Floor Drain Filter Holding Pump Cooler OOE109, and the Reactor Building Equipment Drain Sump Cooler 2(3)E036 during normal plant operation and during a LOOP. (3.1.1) (3.3.1.1.1)

(3.3.1.1.2)

(3.3.1.2.1)

(3.3.1.3.1)

(3.3.1.3.2)

(3.3.1.5.1)

(3.3.2.1.1)

(3.3.2.1.2)

(3.3.4.5) 2.2.1.2.8 Reactor Water Cleanup System (6.1.13.10) 9 The RBCCW System shall suDport operation of the RWCU System by'providing cooling water as required to the RWCU-TCXIIP MOLOl c oolers , the Cle-aTui ToTFRegenerative Heat Ex-changers, and the Cleanup Regenerative Heat

/ Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD NO. P-S-33 Revision 8 Page 19 of 69 3.0 SYSTEM DESIGN BASELINE Section 3.0 provides the system design baseline.

The system design baseline identifies how the system fulfills the design inputs identified in section 2. For every specific design input in section 2.2.1, 2.3.2, 2.4, and 2.5 there is a corresponding discussion in section 3. All discussions in section 3 are traceable to the associated section 2 design inputs via internal DBD referencing. Section 3.1 identifies the system functions and the associated alignments for these functions.

The bases for the system functions are provided via internal "in the text" references to section 2 design inputs.

Section 3.2 identifies the controlling parameters (numerical values) associated with each of the system functions. This includes controlling parameters associated with subsystems and major pieces of equipment. For each controlling parameter, the controlling source document (with external DBD references) and the associated design margin are identified.

Section 3.3 identifies the remaining design features associated with the system. This includes design features associated with subsystems and major pieces of equipment.

The bases for these design features are provided via internal "in the text" references to section 2 design inputs. 3.1 SYSTEM FUNCTIONS Section 3.1 identifies one system function for the RBCCW system: Cooling. The RBCCW System provides the non-safety related function of providing cooling water to the components listed below during all normal plant operating conditions related to power generation and during a LOOP. The following components are cooled by the RBCCW Sys tem : - RWCU Non-Regenerative Heat Exchangers (during - RWCU Pump Motor Coolers (4.7) normal operation only)

Peach Bottom Atomic Power Station, Units 2 and 3 Reactor Building Closed Cooling Water System DBD NO. P-S-33 Revision 8 Page 36 of 69 I Course/Program:

Initial Licensed Operator Training Module/LP ID: PLOT-5003A Title: @ Control Rod Drive Hydraulic Course Code:

ILT Author: S. M. McCartney RevisiodDate:

005 Prerequisites:

N/A Revision By: SMM OPEX Included:

Internal / External / Both / None Est. Duration:

4/50 Minute Periods System (circle one) OBJECTIVES PURPOSE/TERMINAL OBJECTIVES: Familiarize the license trainee with the Control Rod Drive System function, components, operational aspects and their effect on safe facility operation.

TABLE OF CONTENTS (Optional)

Upon successful completion of this lesson, the trainee will be able to: Pg. # 1. Pg. # Describe the relationships between the Control Rod Drive Hydraulic System (CRDH) and the following systems: a. Condensate System

b. Condensate Storage Tanks c. d. e. Reactor Protection System

f. Reactor Manual Control System g. Plant Air Systems h. Control Rod Drive Mechanisms

i. Reactor Water Cleanup Pumps

j. Reactor Pressure Vessel Instrumentation Condensing Chamber Backfill System Reactor Recirculation Pumps (seal purge) Component Cooling Water Systems (TBCCW and RBCCW) 0 Copyright 2000 by Exelon Nuclear, All Riahts Reserved. Permission for reproduction and use is reserved for Exelon Nuclear. (Any other use or reproduction is expressly prohibited without the express permission of Exelon Nuclear.)

I PLOT5003A Rev005 I ContenVSkills

~ ActivitiedNotes

d. Cooling Water Pressure Control Valve (MO-22)

Open-Close, spring return to neutral. Stop button for throttling, however valve is left wide open.

e. Scram Discharge Volume Vent and Drain Valves 1) Two handswitches: Each switch operates 3 valves. Each switch can block off vent and drain paths. 2) Open-Close, spring returns to auto. f. Stabilizing Valve Control 1) Can select in control room which set receives control signal from RMCS system. 2) Desired set of stabilizing valves must be manually valved in. 4. Interlocks

a. CRDPump Pump will trip on low suction pressure and various electrical malfunctions.

b. Scram Discharge Volume 1) Rod Block 2) Scram 3) Can be bypassed by SDV High Volume Scram bypass switch E. System Operation

1. Systems Interrelations

a. RMCS supplies control power to directional control valves and stabilizing valves. b. RPS controls the operation of the scram pilot valves, scram valves, backup scram pilot valve, and SDV vent and drain valves. c. CRDH supplies Reactor Recirc pumps with seal purge water.

1 PLUT5003A Rev005 Page 18 of 24

~~ ActivitiedNotes Content/Skills

3. Effects of the loss of other systems on the CRDH system a. The loss of the condensate header supply will cause the CRD pump to draw from the CST. If both CST and condensate are lost the CRD pump will trip on low suction pressure.

b. The loss of instrument air will cause a scram. This has the same effect as a normal scram. The FCVs (AO-lgA, 6) fail shut on a loss of plant air. c. The loss of the RPS will cause a scram to occur since the scram pilot valves deenergize.

d. A loss of AC power to the CRD pumps will cause them to trip. The loss of AC power to RPS will cause a scram. e. A loss of TBCCW and RBCCW will cause the CRD pump to overheat.

F. Technical Specifications Using the current revision of Technical Specifications and Bases, discuss the following for each of the listed Specifications:

0 0 LCO and Applicability 0 ACTIONS SRs and implementing Operations STs 1. TS 3.1.3 Control Rod OPERABILITY

2. TS 3.1.4 Control Rod Scram Times 3. TS 3.1.5 Control Rod Scram Accumulators

4. TS 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves Provide exercises to apply TSAs. I PLOT5003A Rev005 Page 20 of 24 RM Pox crmna m1 UMWOERS I 1 I I I 4 3 7 6 5 t 8 CATEGORY AI 19E -W-09291 I 8 -& r 1 I I I I t I CATEGORY AI AO 10.4-2 Rev. 16 Page 1 of 20 MTW: mtw Exelon Nuclear Peach Bottom Unit 2 A0 10.4-2 RESIDUAL HEAT REMOVAL SYSTEM - FUEL POOL TO REACTOR MODE 1.0 2.0 PURPOSE This procedure provides the instructions necessary for placing an RHR Pump and Heat Exchanger in service in the Fuel Pool to Reactor mode. PREREQUISITES 2.1 2.2 2.3 2.4 2.5 2.6 2.7 2.8 2.9 4KV power available to the RHR System in accordance with so 54. 480 VAC MCC power available to the RHR System in accordance with SO 56E. 250 VDC power available to the RHR System in accordance with SO 57B.1-2, "125/250 Volt Station Battery Charger Operationsll . Verify RHR pump power will not be supplied from a diesel generator. Instrument Air System available to the RHR System in accordance with SO 36B. Fuel pool gates to reactor cavity removed.

Reactor in MODE 5. HP notified for potential dose rate changes in Reactor Building Elev. 165' due to flow through the RHR/FP Spool piece. Verify the Fuel Pool Skimmer Surge Tank Level on LI-2695 is able to be displayed in the Main Control Room in view of the Reactor Operator.

2.10 Verify RHR Shutdown Cooling is in operation is aligned for Shutdown Cooling operation on the "A" or rrD1l Pump in accordance with SO 10.l.B-2, lrResidual Heat Removal System Shutdown Cooling Mode Manual Start". 2.11 IF Section 4.3 is to be performed, THEN verify AO-2-10-046A(B), "RHR Loop A(B)

Check Valve1! AO-2-10-163A(B), "RHR Recirc Loop A(B) Testable Check Valve Equalizer", are capable of performing its isolation function due to loss of automatic isolation of MO-2-10-025A(B).

A0 10.4-2 Rev. 16 Page 2 of 20 2.12 Fuel Pool to RHR Pump Suction piping flushed in accordance with SO 10.5.A-2, IIResidual Heat Removal System Piping Flush". 3.0 PRECAUTIONS 3.1 Prior to removing a reactor coolant circulation method from service refer to GP-12, IICore Cooling Procedurer1.

During the performance of this procedure, the normal shutdown cooling subsystem will be inoperable per Tech Specs, however, this procedure does provide for reactor coolant circulation. Reference Tech Spec 3.9.7.C. 3.2 During the period this procedure is in effect, the manual restoration of the suction path of shutdown cooling may not be available.

Reference Tech Spec 3.9.7. 3.3 IF reactor temperature limits cannot be maintained per Tech Spec 3.4.9 with this lineup, THEN establish normal shutdown cooling and restore fuel pool cooling.

3.4 WHEN selecting a RHR pump, THEN check the opposite unit AND verify the like pump is NOT in service. 4KV interlocks prevent simultaneous operation of the same pump on the opposite unit. 3.5 One restart of the RHR pumps from rated temperature is permitted; then the motor shall be allowed to cool to ambient temperature before restarting.

Two starts are permitted from ambient temperature every 30 minutes. 3.6 Minimize the amount of time the RHR flow is less than 4,000 gpm. Do NOT operate with steady state flow less than 4,000 gpm to prevent possible pump damage. 3.7 Do NOT operate above a maximum RHR flow of 6,500 gpm due to loss of pump suction AND a loss of makeup to the skimmer surge tank. 3.8 During the performance of this procedure, WHEN venting draining equipment, ensure the capacity of the drain is NOT exceeded.

3.9 The RHR Pump will lose suction if make-up to the Skimmer Surge Tank is lost. If the Skimmer Surge Tank Low Level Alarm annunciates indicating a level of 60", THEN suction to the RHR pump would be lost in approximately 30 seconds. 3.10 IF Skimmer Surge Tank level cannot be maintained above the low level alarm point, THEN Shut Down Cooling (SDC) flow must be immediately reduced to restore level above 60". IF Skimmer Surge Tank level cannot be immediately restored above 60", the RHR Pump shall be secured.

A0 10.4-2 Rev. 16 Page 3 of 20 3.11 RV-2-10-035A(B) may lift on RHR pump starts with reactor level above the RPV flange. Prior to starting any FUR pump with reactor level above the RPV flange, notify HP and verify that personnel are evacuated from the following bays: o I1A1l Loop RHR - Bays 12, 13, 14 lrB1l Loop RHR - Bays 4, 5, 6 3.12 The normal RHR suction path for Shutdown Cooling will be isolated by closing MO-2-10-17 OR MO-2-10-018 OR both, OR by closing HV-2-10-88.

4.0 PERFORMANCE STEPS NOTES 1. Section 4.1, Establishes a suction path from the Fuel Pool

2. Section 4.2, Isolates the RHR Suction from the reactor via Skimmer Surge Tank to the RHR Suction. HV-2-10-88, IIShutdown Cooling Suction From Recirc Loop A Is01 Valve". 3. Section 4.3, Isolates the RHR Suction from the reactor via 4. Section 4.4, Temporarily removes an RHR Pump and Heat Exchanger, 5. Section 4.5, Restores an RHR Pump and Heat Exchanger to service MO-2-10-017 OR MO-2-10-018 OR both. in the Fuel Pool to Reactor Mode, from service. in the Fuel Pool to Reactor Mode after temporary removal. Mode. 6. Section 4.6, Securing RHR operating in the Fuel Pool to Reactor 4.1 Establish Fuel Pool Skimmer Surge Tank to RHR suction path with normal RHR Shutdown Cooling suction path aligned to the reactor. 4.1.1 Verify the RHR/FP Cross Tie spool piece which ties the fuel pool skimmer surge tanks to the RHR System is installed at Rx Bldg, 165' El. 4.1.2 Notify the Fuel Handling Director Reactor Engineering that the Fuel Pool Cooling System may be removed from service to support this A0 procedure AND evaluate the impact on fuel floor activities per FH-GC, "Core Component Movement-Core Transfer", prerequisites.

A0 10.4-2 Rev. 16 Page 4 of 20 4.1.3 4.1.4 4.1.5 4.1.6 IF required, verify Fuel Pool Cooling is secured OR secure Fuel Pool Cooling in accordance with SO 19.2.A-2, IIFuel Pool Cooling System Component Removal and System Shutdown".

N/A if not required.

Verify two fuel pool to skimmer surge tank weir gates and two reactor cavity to skimmer surge tank weir gates are in their lowest position, direct NMD to lower two fuel pool to skimmer surge tank weir gates and two reactor cavity to skimmer surge tank weir gates to their lowest position.

IF RHR Shutdown Cooling is operating, THEN throttle CV-2-10-2677A(D) flow rate between 4,000 gpm and 4500 gpm.

IF RHR Shutdown Cooling is NOT operating, THEN perform the following.

Otherwise, N/A these steps. 4.1.6.1 Verify the A/C Selector switch for CV-2-10-2677A(D) is rrOFF1l on Panel 20C716 (20C717). . 4.1.6.2 Throttle CV-2-10-2677A(D) ten handwheel turns open from full closed. CAUTION Unisolating the Fuel Pool to RHR suction path in Steps 4.1.8 through 4.1.10 will make RHR Shutdown Cooling inoperable per Tech Spec 3.9.7.A 3.9.7.C. This procedure does provide for reactor coolant recirculation. Reference Tech Spec 3.9.7.A. 4.1.7 Prior to performing Steps 4.1.8 through 4.1.10, commence performing ST-0-080-500-2, "Recording and Monitoring Reactor Vessel Temperatures and Pressuret1 to ensure compliance with Tech Spec Action 3.9.7.A and 3.9.7.C, as required. 4.1.8 Direct an operator to unlock AND slowly open HV-2-19-25, "Surge Tanks to RHR System Valve".

SO 19.1.A-2 Rev. 15 Page 1 of 16 MDF : mdf PECO Energy Company Peach Bottom Unit 2 SO 19.1.A-2 FUEL POOL COOLING SYSTEM STARTUP AND NORMAL OPERATIONS (This revision is a total rewrite) 1.0 PURPOSE This procedure provides instructions necessary to establish flow in the Fuel Pool Cooling System for the removal of decay heat from the Spent Fuel Pool. This procedure also provides instructions to place additional Fuel Pool Cooling components in service as required.

2.0 PREREQUISITES 2.1 See individual sections.

3.0 PRECAUTIONS 3.1 The amount of pumps and heat exchangers required to maintain the Spent Fuel Pool temperature from exceeding the maximum of 130°F will vary with system heat load.

3.2 The Fuel Pool F/D should be removed from service for regeneration when F/D delta pressure exceeds 25 psid. 3.3 Rapid flow adjustments may cause severe water hammer. 3.4 Mispositioning of the chain operated valves may result in cross-tying the U/2 and U/3 Fuel Pools. 3.4.1 It is essential to check the desired direction of valve stroke when pulling the chain, because the valves are located above and behind the operator manipulating the chain. 3.4.2 These valves are llKnockertf type valves which require the operator to knock the valves free in the desired direction.

3.5 This procedure is NOT to be used for placing the tfC1l Demin in-service on Unit 2. For this configuration, SO 19A.7.D-2, "Placing Additional Fuel Pool Filter Demineralizers in Service and Removal of the rrC1l Demineralizers From Service when Aligned to the Unit 2 Fuel should be referenced.

3.6 Normal alignment of in service Fuel Pool Cooling System components is comprised of one Fuel Pool Service Water Booster Pump and one Fuel Pool Cooling Pump per one Fuel Pool Cooling Water Heat Exchanger.

The number of in-service Fuel Pool Cooling Water Pumps should NOT be greater than the number of in-service heat exchangers.

SO 19.1.A-2 Rev. 15 Page 6 of 16 NOTES Attachment 1 provides details on operating the Moore controllers FCS-0-19-4-069A(B)

IIFuel Pool F/D Outlet A(B)

Flow". The fuel pool cooling pump discharge pressure allowable operating band is 105 to 115 psig. - IF maximum Fuel Pool Cooling is desired, THEN the operating Fuel Pool Cooling Pump discharge pressure should be adjusted to the low end of the operating band.

Otherwise, the operating Fuel Pool Cooling Pump discharge pressure should be adjusted to 110 to 115 psig. Fuel Pool Cooling Pump Discharge Pressure of 105 to 115 psig as read on PI-2703A,BtC on Panel 20C076 should be maintained by performing Steps 4.1.15.4 4.1.15.5 concurrently.

4.1.15.4 Slowly adjust controller FCS-0-19-4-069At "Fuel Pool F/D Outlet A Flow" to raise flow, with flow NOT to exceed 550 gpm, then place filter demin flow controller to "AUTO" if NOT in IIAUTO1l.

Flow Controller may be left in IIMANUALII if "AUTOvv is unstable.

& 4.1.15.5 Throttle HV-2-19-46, "Fuel Pool Filter Demin Bypass Valvet1, as required to maintain Fuel Pool Cooling discharge pressure in the required band for plant conditions at PI-2703A(BJC) at Panel 20C076. 4.1.15.6 Perform the following at Panel OOC110. o Place filter demin hold pump to "AUTO . o Place AO-0-19-23A, "Hold Pp Disch Valve H," control switch to "AUTO". 4.1.15.7 Proceed to Step 4.1.17 4.1.16 Place the IrBII Filter Demin in service on Unit 2. 4.1.16.1 Verify the IIB" Fuel Pool F/D is NOT aligned to Unit 3 that the lIB" F/D can be aligned to Unit 2. 4.1.16.2 Verify the filter demin is in standby in accordance with SO 19A.7.B, "Fuel Pool Cooling Filter Demineralizer Automatic Regeneration1I , SO 19A. 7. C, "Fuel Pool Cooling Filter Demineralizer Manual Regeneration".

__ - ~~ - ~ Exelon Nuclear Log Query Output Page 1 of 30 There were 258 matches to your query which was: Peach Bottom's Archival Operations Narrative Logs where the Logdates were after 09/20/2003 12:OO:OO AM and before 09/21/2003 11:59:59 PM 0 Back to Selection Page Note: EXELON INTERNAL USE ONLY - The data displayed in these reports is not an official record and is not intended to be used for decisions affecting plant operation.

ISN'IKY procedure SO 20C.7.L PROCESSING THE WASTE TANK. Status: satisfactorily Detail:

wcf off on high d/p. wst level 34%. waste surge tank level 74% 9/20/2003 12:29:31 AM 9/20/2003 12:29:40 AM 9/20/2003 12:42:3 8 AM 9/20/2003 1:00:58 AM ENTERED AN UNMET REGULATORY ACTION Item Number: 03-3-131 Affected Unit: 3 Entry Type (TSA, PTSA, TRM, FTRM, ODCM, PODCM): PTSA Specification Entered: 3.5.2 Entered Datemime: 9/20/03 0827 Equipment ID: GP-20 System Number: 10,14 Reference Number(s):

GP-20 Condition (s) Entered: None Is a SFD required? (TSA entries ONLY) N/A Are any other SFDs currently active? (TSA entries ONLY)

N/A Reason(s) Entered: Unit 3 ECCS auto initiation defeated iaw GP-20. Required Compensatory Action(s) or Limitation(s): Maintain reactor level

> 458" with the Fuel Pool gates removed and No OPDRVs inprogress or comply with Tech Spec 3.5.2 Limiting Completion Datemime:

N/A Required Compensatory Action(s) or Limitation(s):

N/A Limiting Completion Datemime:

N/A Required Compensatory Action( s) or Limitation(s): N/A Limiting Completion Dateflime:

N/A Required Compensatory Action(s) or Limitation( s): N/A Limiting Completion Datemime:

N/A Required Compensatory Action(s) or Limitation(s):

N/A Limiting Completion Datemime:

N/A Entered By: Breidenbaugh Verified By: Pautler Performed procedure ST-0-60F- 100-2 FUNCTIONAL TEST OF RPS CHANNEL A SCRAM TEST SWITCHES. Status: satisfactorily Detail: FUNCTIONAL TEST OF RPS CHANNEL A SCRAM TEST SWITCHES COMPLETED. Entered Procedure SO 20A. 1 .D FLOOR DRAIN COLLECTOF TANK NORMAL PROCESSING TO FLOOR DRAIN SAMPLE TANK. Detail: placed the fdf i/s to the fdct.initia1 fdct level 44%, initial fdst level 64%. filter # 092001-1 ENTERED AN UNMET REGULATORY ACTION Item Number: 03-3-132 Affected Unit: 3 Entry Type (TSA, PTSA, TRM, PTRM, ODCM, PODCM): FTSA Specification Entered:

PTSA 3.1.4 Entered Datemime: 9/20/2003 1:00:58 AM ;s 1: K'l Y PI.: tW J-3 J-2 <W J-0 iOOOlch i00Odsp http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&narr. . . 7/11/2005 Page 27 of 30 Exelon Nuclear Log Query Output I I because the surge tanks are crosstied. satisfactorily Detail: transfer complete. wct level 24%.

surge tank level 28% SAMPLE TANK. Detail: placed the fdf i/s to the fdst. fdct level 34% fdst level 26%.

filter## 0921 10-1 FROM COLLECTOR TANK TO SAMPLE TANK. Detail: placed the wcf i/s to 'b' wst. wct level 24%,wst level 5%. filter #l RHR SDC MWE: zero zero MANUAL START. Status: satisfactorily Detail: Placed "A" RHR pump in service in SDC Mode after temporary shutdown unsatisfactorily Detail:

TEST WAS COMPLETED UNSAT DUE TO VENT STACK RAD MONITOR RI-2979B BEING INOPERABLE. REFERENCE A1401 829 AND TRM-03 142 OPERATIONS. Status: satisfactorily Detail: placed 2K condensate demin in service. Status: satisfactorily Detail: Removed 2G condensate demin from service. Detail: Commenced regen of 2G condensate demin. Closed indication failed to light for the 2G 'A' valve (A1435196).

The precoat outlet valve failed to open, causing precoat tank level to rise (A1433615) 1W 1W 1W J-0 J-3 J-0 J-2 (LO-2 IJLO-2 ao-2 J-3 http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&narr. . . iOOOlch iOOOlch iOOOlch iOO2cjc i00Otbm iOO4dak i003dlh iOOOstr 100ostr 1OOOStr iOOOtbm 7/11/2005 Exelon Nuclear Log Query Output 9/2 1/2003 11:33:33 ~PM 1 Performed procedure ARC-3 17 30C2 12R H-2 "D" DRYWELL COOLER AIR HI TEMP. Status: satisfactorily Detail: PREVIOSLY AR'D A1415681.

SET POINT 135 DEG F, CURRENT READING 65 DEG F. AR UPDATED. 9/21/2003 115657 PM Performed procedure SO 50A.7.D-2 MAINTAINING STATOR COOLING WATER STORAGE TANK LEVEL. Status: satisfactorily Detail: Filled U/2 Stator cooling tank 1/2 in the sightglass per the proc. SUSPENDED FUEL MOVEMENT (SHUFFLE

1) DUE GRAPPLE MALFUNCTION. Back to Selection Page 9/2 1/2003 11:59:00 PM Page 29 of 30 Prompt investigation initiated iaw OP-AA- 106- 101- 1001 for CR 176768 due to not having the U/2 MSIV's opened iaw GP- 2. All OP-AA-106-101 notifications have been comdeted.

IW (LO-3 J-3 LW JLO-0 J-2 J-3 (LO-2 J-3 00Olch 100 1 kpp OOOtbm iO00lch Io00gwp 1003dlh 6OOtbm IOOldja iO00tbm iOOOrj f http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&narr.

.. 7/11/2005

~ __ - - ~~ Exelon Nuclear Log Query Output 1/22/2003 2:41: 10 1 Page 1 of 35 Performed procedure SO 38C.2.A MAKEUP WATER SYSTEM SHUTDOWN. Status: satisfactorily Detail: CWST There were 288 matches to your query which was: Peach Bottom's Archival Operations Narrative Logs where the Logdates were after 09/22/2003 12:OO:OO AM and before 09/23/2003 11:59:59 PM 0 Back to Selection Page Note: EXELON INTERNAL USE ONLY - The data displayed in these reports is not an official record and is not intended to be used for decisions affecting plant operation. .O( ;I)'.irll<

I EN'I'KY U/2, LOWERING RX PRESSURE BY INSERTING CONTROL RODS TO LOWER DIFFERENTIAL PRESSURE BETWEEN MAIN STEAM LINES AND REACTOR PRESS. THIS IS NECESSARY TO OPEN MAIN STEAM ISOLATION VALVE

S. PROCEDURE

CONTROL OF THIS EVOLUTION IS GP-2. RODS WERE INSERTED FROM STEP 100 TO STEP 80 IN GP-2 TO STOP REACTOR HEATUP. REACTOR WAS TAKEN SUBCRITICAL AND WILL REMAIN SUBCRITICAL UNTIL MSIV'S ARE OPEN CR #176768. INITIATION. Status: satisfactorily Detail: PERFORMED RCIC SYSTEM ALIGNMENT FOR AUTO OR MANUAL FROM COLLECTOR TANK TO SAMPLE TANK. Status:

satisfactorily Detail: wcf off on high d/p and high wst level of 98%. wct level is 34% J-2 !W IW JLO-2 J-2 4LO-0 AI, CIS *3 ~003dlh ~003dlh iOOOlch kOlch iOOOstr i003dlh l00ogwp http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&narr.

.. 7/11/2005

~~ Exelon Nuclear Log Query Output Page 14 of 35 1/23/2003

08:52 AM FIRE BRIGADE DISPATCHED. ALARM WAS DUE TO GRINDING IN THE DRYWELL. WORK WAS STOPPED AND FIRE WATCH REMAINED. MAIN CONTROL ROOM ALARM RESET. Performed procedure GP-2 NORMAL PLANT START-UP.

Status: satisfactorily Detail: STARTED RAISING REACTOR PRESSURE FROM 450 PSIG TO 940 PSIG. ENTERED AN UNMET REGULATORY ACTION Item Number: 03-2-154 Affected Unit: 2 Entry Type (TSA, FTSA, TRM, PTRM, ODCM, PODCM): FTSA Specification Entered:

n/a Entered Datemime:

n/a Equipment ID: n/a System Number: 60F Reference Number(s):

GP-2 Condition(s) Entered: none Is a SFD required? (TSA entries ONLY) n/a Are any other SFDs currently active? (TSA entries ONLY) n/a Reason(s) Entered: The TCVmSV fast closure scram is bypassed IAW GP-2 attachment

7. Operation greater than or equal to 29.5% thermal power is not permitted. Required Compensatory Action(s) or Limitation(s): Maintain core thermal power

< or = 29.5% or comply with TS 3.3.1.1 and 3.3.4.2 Limiting Completion Datemime: Required Compensatory Action( s) or Limitation(s): Limiting Completion Datemime: Required Compensatory Action(s) or Limitation(s): Limiting Completion Datemime: Required Compensatory Action(s) or Limitation(s): Limiting Completion Datemime: Required Compensatory Action(s) or Limitation(s): Limiting Completion Datemime:

Entered By: R Llewellyn Verified By:

R Glackin MODE MANUAL START. Status: satisfactorily Detail: 3A RHR PUMP SHUTDOWN TEMPORARILY FOR FUEL POOL CLARITY. SHUTDOWN COOLING IS OUT OF SERVICE. MANUAL START. Status: satisfactorily Detail:

3A RHR PUMP RESTARTED. SHUTDOWN COOLING HAS BEEN u-2 u-2 u-3 RW u-3 u-3 u-2 u003dlh u002rsl uOOOtbm uOOOlch uOOOtbm uOOOtbm u003dlh http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&narr. . . 7/11/2005

___~___ ~ Exelon Nuclear Log Query Output Page 1 of 37 There were 385 matches to your query which was:

Peach Bottom's Archival Operations Narrative Logs where the Logdates were after 09/15/2004 12:OO:OO AM and before 09/17/2004 11:59:59 PM 0 Back to Selection Page Note: EXELON INTERNAL, USE ONLY - The data displayed in these reports is not an official record and is not intended to be used for decisions affecting plant operation.

I.OGl)A?'li~

HN'I'KY Start SO 20A.l.D XFR FDCT => FDST Filter

1. 1. 091501-1 Initial FDCT Level:

30 % Initial FDST Level:

82 96 Comments: DRAIN PIPING HOT SPOT FLUSH. Status: satisfactorily Detail: Performed flush per recommendation of HP's to reduce CONTAINMENT REASON - PRIMARY CONTAINMENT INOPERABLE. Must be restored to operable prior to entry to WHO IS EXHIBITING SYMPTOMS OF HEAT STRESS.

DISPATCHED INCIDENT COMMANDER AND OPERATIONS HEALTH PHYSICIST IN ADDITION TO EL JARKOWSKI FROM FOURTH FLOOR ADMIN BUILDING. INDIVIDUAL'S NAME IS BRIAN RADWASTE BUILDING 135' EL TO COOL DOWN.

NO FURTHER ASSISTANCE IS REQUIRED.

GUDERYON.

SSN 399-70-53

17. INDIVIDUAL MOVED TO ump down complete. Final level 14.65' Broke vacuum on U/2 for outage activities CSE K'I k' Plii <W J-2 J-2 5-2 J-0 J-2 J-2 J-2 41,USEH 002efh OOlbsb 00 1 bsb ,005dlf 1004kms 100 1 bsb 100 1 bsb 100 1 bsb http://opt.exeloncor.com/getvar.asp?nulli~e~&~chivehid=&subloghid=&sitehid=&n~at

... 7/7/2005 Fxelon Nuclear Log Query Output - L - PLACED THE WCF I/S =>'B' WST(LVL

@ 32%) @ 80GPM. W91504-5.DP

@ 4#. WCT @ 64%. procedure SO 10.l.B-2 RESIDUAL HEAT MANUAL START. Status: satisfactorily Detail: 2D RHR ump placed in service in shut down cooling.

SO 40C.7.A-2 PRIMARY Status: satisfactorily Detail: Removed DW purge from service IAW section 4.8 and started torus purge IAW section 4.5. WATER Detail: removed the ionics skid from service; final dst level @ 20 feet. Performed procedure SO 19.7.E-2 ALIGNING FUEL POOL COOLING SYSTEM TO REACTOR WELL. Status: satisfactorily Detail: 2A

& 2B pumps, heat exchangers and demins are in service. Reactor Power

@ 100% IAW SO 2A. 1 .D- 3 and GP-5. Initial speed: 1477-1491 RPM. Final speed: 1485-1497 RPM.

9/16/2004 Hap McDaniel relieved Adam Buckley as the unit 2 reactor 1:08:45 PM operator.

9/16/2004 Entered Procedure SO 20A.7.N TRANSFER FLOOR DRAIN 1 :30:47 PM SURGE TANK TO WASTE SURGE TANK.

Detail: DRAIN SURGE TANK TO THE WASTE SURGE TANK. 9/16/2004 1 1 :35:23 RAISED the speed of the U/3 "A" Recirc Pump to maintain OPENED HV-2-20C- 11429 TO CROSS TIE THE FLOOR procedure SO 40C.7.A-2 PRIMARY VENTILATION.

Status: satisfactorily Detail: Secured Torus purge IAW section 4.8 and placed DW urge in service IAW section 4.4 . H SO 28A.5.A-2 OPERATION OF BETZ- satisfactorily Detail: swapped condenser cleaning from the "3B" to the "3C" main condenser. 108- 1 1 1 ADVERSE PLANNING.

Status: satisfactorily Detail: RV-7 1A TAILPIPE TEMPERATURE IS 257.8 DEGF (TR-3-02-103 POINT

1. 3) OPERATIONS.

Status: satisfactorily Detail: TRANSFER U/3 CBRT(LVL 64%=>20%)

TO THE 3A CPS(LVL 34%=>99%)

19/16/2004ll~~~~~

SLOAN TEMPORARILY RELIEVED MIKE AMES Page 22 of 37 1-2 1-2 LO-0 J-2 J-3 J-2 J-2 ILO-0 J-3 J-2 IW J-0 ~005dab 1005dab 1005dab 1000dkh i005dab rO00hrp i005dab 1004dlh iO00dkh 1005dab iOOOhrp i00Omla http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&na~at

... 7/7/2005 Exelon Nuclear Log Query Output 9/17/2004 4:30:00 AM Page 29 of 37 Performed procedure SO 5A.6.A-3 PLACING STANDBY CONDENSATE DEMINS IN SERVICE & NORMAL OPERATIONS. Status: satisfactorily Detail: RETURNED 3K C/D TO SERVICE. II 11% Comments: - EXITED AN UNMET REGULATORY ACTION Item Condition(s) Exited: TRM 3.14.8 CONDITIO 'A' Reasons Exited: DOOR #190 RETURNED TO OPERABLE STATUS AS A FIRE BARRIER WITH COMPLETION OF WORK BBP 04-250. Closing Action(s) Entered By: P. PAUTLER Closing Action(s)

Verified By: D. FORRY Number: 04-2-027 Entry Type (TSA, PTSA, etc.): TRM UNDER CO211250-01 TO REPLACE DOOR STRIKE. REF Status: satisfactorily Detail: FILLED DOMESTIC WATER HYPOCHLORITE INJECTION MIX TANK (SECTION 4.3). Level Feet Comments: SURGE TANK THRU THE WCF. INITIAL WASTE SURGE TANK LVL=60%, INITIAL 'A' WST LVL= 32%. satisfactorily Detail:

DST @ 17', CWST 0 24'. Status: satisfactorily Detail: SDC returned to service through the 2D RHR pump at 5000gpm. Status: satisfactorilv Detail: REGENED U/3 "K" C/D. to Condensate Phase Sep:

2A Checkbox - YES 2B Checkbox - NO 3A Checkbox - NO 3B Checkbox - NO Initial CPS Level is complete 19', CWST 63 23'. http://opt.exeloncorp.com/getvar.asp?nullifier=&archivehid=&subloghid=&sitehid=&narrat

... 7/7/2005